Text

Waterford Steam Electric Station, Unit 3 - Issuance of Amendment No. 249 Adoption of TSTF-425, Relocate Surveillance Frequencies to Licensee Control (CAC No. MF6366)
	ML16159A419
Person / Time
Site:	Waterford
Issue date:	07/26/2016
From:	Pulvirenti A L; Plant Licensing Branch IV
To:	; Entergy Operations
	Pulvirenti A L
References
	CAC MF6366
	Download: ML16159A419 (144)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 26, 2016 Site Vice President Entergy Operations, Inc. Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 -ISSUANCE OF AMENDMENT RE: ADOPTION OF TSTF-425, REVISION 3 "RELOCATE SURVEILLANCE FREQUENCIES TO LICENSEE CONTROL -RITSTF INITIATIVE 5b" (CAC NO. MF6366)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 249 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated June 17, 2015, as supplemented by letters dated March 3, April 28, and July 12, 2016. The amendment modifies the TSs by relocating specific surveillance frequencies to a licensee-controlled program. The proposed changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF [Risk-Informed TSTF] Initiative 5b." A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Docket No. 50-382

Enclosures:

1. Amendment No. 249 to NPF-38 2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, April L. Pulvirenti, Project Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS. INC. DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 249 License No. NPF-38 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Entergy Operations, Inc. (EOI), dated June 17, 2015, as supplemented by letters dated March 3, April 28, and July 12, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations 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. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. 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. Enclosure 1 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. NPF-38 is hereby amended to read as follows: 2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 249, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.

Attachment:

Changes to the Facility Operating License No. NPF-38 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION f /,£__ Acting Chief Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: July 26, 201 6 ATTACHMENT TO LICENSE AMENDMENT NO. 249 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Facility Operating License and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT REMOVE INSERT REMOVE INSERT REMOVE INSERT 1 1-7 3/4 2-12 3/4 2-12 3/4 4-9a 3/4 4-9a 3/4 6-40 3/4 6-40 ' 1-9 1-9 3/4 2-13 3/42-13 3/4 4-17a 3/4 4-17a 3/4 7-5 3/4 7-5 3/4 1-1 3/4 1-1 3/4 3-1 3/4 3-1 3/4 4-18 3/4 4-18 3/4 7-6 3/4 7-6 3/4 1-2 3/4 1-2 3/4 3-2 3/4 3-2 3/4 4-19 3/4 4-19 3/4 7-8 3/4 7-8 3/4 1-3 3/4 1-3 3/4 3-10 3/4 3-10 3/4 4-25 3/4 4-25 3/4 7-9 3/4 7-9 3/4 1-5 3/4 1-5 3/4 3-11 3/4 3-11 3/4 4-29 3/4 4-29 3/4 7-9a 3/4 7-9a 3/4 1-6 3/4 1-6 3/4 3-12a 3/4 3-12a 3/4 4-35 3/4 4-35 3/4 7-9b 3/4 7-9b 3/4 1-7 3/4 1-7 3/4 3-13 3/4 3-13 3/4 4-37 3/4 4-37 ---* 3/4 7-9c 3/4 1-9 3/4 1-9 3/4 3-25 3/4 3-25 3/4 5-2 3/4 5-2 3/4 7-11 3/4 7-11 3/41-11 3/4 1-11 3/4 3-26 3/4 3-26 3/4 5-2a 3/4 5-2a 3/47-13 3/47-13 3/4 1-12 3/4 1-12 3/4 3-27 3/4 3-27 3/4 5-4 3/4 5-4 3/4 7-16a 3/47-16a 3/4 1-13 3/4 1-13 3/4 3-32 3/4 3-32 3/4 5-5 3/4 5-5 3/47-17 3/4 7-17 3/4 1-16 3/4 1-16 3/4 3-33 3/4 3-33 3/4 5-9 3/4 5-9 3/47-18a 3/4 7-18a 3/4 1-20 3/4 1-20 3/4 3-43 3/4 3-43 3/4 6-1 3/4 6-1 3/4 7-19 3/4 7-19 3/4 1-21 3/4 1-21 3/4 3-46 3/4 3-46 3/4 6-10 3/4 6-10 3/4 7-20 3/4 7-20 3/4 1-22 3/4 1-22 3/4 3-47 3/4 3-47 3/46-11 3/46-11 3/4 7-43 3/4 7-43 3/4 1-24 3/4 1-24 3/4 3-48a 3/4 3-48a 3/4 6-13 3/4 6-13 3/4 8-3 3/4 8-3 3/4 1-26 3/4 1-26 3/4 3-65 3/4 3-65 3/4 6-15 3/4 6-15 3/4 8-4 3/4 8-4 3/4 2-1a 3/42-1a 3/4 4-1 3/4 4-1 3/4 6-16 3/4 6-16 3/4 8-5 3/4 8-5 3/4 2-3 3/4 2-3 3/4 4-2 3/4 4-2 3/4 6-17 3/4 6-17 3/4 8-6a 3/4 8-6a 3/4 2-5 3/4 2-5 3/4 4-4 3/4 4-4 3/4 6-18 3/4 6-18 3/4 8-8a 3/4 8-8a 3/4 2-6a 3/4 2-6a 3/4 4-5a 3/4 4-5a 3/4 6-20 3/4 6-20 3/4 8-9 3/4 8-9 3/4 2-10 3/42-10 3/4 4-6 3/4 4-6 3/4 6-37 3/4 6-37 3/4 8-10 3/4 8-10 3/4 2-11 3/4 2-11 3/4 4-9 3/4 4-9 3/4 6-38 3/4 6-38 3/4 8-14 3/4 8-14 314 8-15 314 8-15 314 9-4 314 9-4 314 9-13a 314 9-13a 6-9 6-9 314 8-16 314 8-16 314 9-8 314 9-8 314 10-1 314 10-1 314 8-17 314 8-17 314 9-9 314 9-9 314 10-2 314 10-2 314 8-52 314 8-52 314 9-11 314 9-11 314 10-3 314 10-3 3/4 9-1 314 9-1 314 9-12 314 9-12 314 10-5 314 10-5 314 9-2 314 9-2 314 9-13 314 9-13 3/411-17 3/411-17 or indirectly any control over (i) the facility, (ii) power or energy produced by the facility, or (iii) the licensees of the facility. Further, any rights acquired under this authorization may be exercised only in compliance with and subject to the requirements and restrictions of this operating license, the Atomic Energy Act of 1954, as amended, and the NRC's regulations. For purposes of this condition, the limitations of 10 CFR 50.81, as now in effect and as they may be subsequently amended, are fully applicable to the equity investors and any successors in interest to the equity investors, as long as the license for the facility remains in effect. (b) Entergy Louisiana, LLC (or its designee) to notify the NRC in writing prior to any change in (i) the terms or conditions of any lease agreements executed as part of the above authorized financial transactions, (ii) any facility operating agreement involving a licensee that is in effect now or will be in effect in the future, or (iii) the existing property insurance coverages for the facility, that would materially alter the representations and conditions, set forth in the staff's Safety Evaluation enclosed to the NRC letter dated September 18, 1989. In addition, Entergy Louisiana, LLC or its designee is required to notify the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on the operation of the facility. C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I 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 EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein. 2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 249, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. AMENDMENT NO. 249 DEFINITIONS SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land or property is not owned, leased, or otherwise controlled by the licensee. SOF1WARE 1.30 The digital computer SOF1WARE for the reactor protection system shall be the program codes including their associated data, documentation, and procedures. 1.31 Definition 1.31 has been deleted. SOURCE CHECK 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. 1.33 Definition 1.33 has been deleted. THERMAL POWER 1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. UNIDENTIFIED LEAKAGE 1.35 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. WATERFORD-UNIT 3 1-7 Amendment No. 68, 116, 249 NOTATION s D w M p Q SA R S/U N.A. SFCP WATERFORD -UNIT 3 TABLE 1.1 FREQUENCY NOTATION FREQUENCY At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . At least once per 7 days. At least once per 31 days. Completed prior to each release. At least once per 92 days. At least once per 184 days. At least once per 18 months. Prior to each reactor startup. Not applicable. Surveillance Frequency Control Program 1-9 AMENDMENT NO. 249 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORA TION CONTROL SHUTDOWN MARGIN -ANY CEA WITHDRAWN LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to that specified in the COLR. APPLICABILITY: MODES 1, 2*, 3, 4, and 5 with any CEA fully or partially withdrawn. ACTION: With the SHUTDOWN MARGIN less than that specified in the COLR, immediately initiate boration to restore SHUTDOWN MARGIN to within limit. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 With any CEA fully or partially withdrawn, the SHUTDOWN MARGIN shall be determined to be greater than or equal to that specified in the COLR:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable CEA(s). b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, by verifying that CEA group withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6 in accordance with the Surveillance Frequency Control Program. c. When in MODE 2 with Keff less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6. See Special Test Exception 3.10.1. WATERFORD -UNIT 3 314 1-1 AMENDMENT NO 11, 33, 102, 141, 182, 249 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6. e. When in MODE 3, 4, or 5, in accordance with the Surveillance Frequency Control Program by consideration of at least the following factors: 1. Reactor Coolant System boron concentration, 2. CEA position, 3. Reactor Coolant System average temperature, 4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration. 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1.0% delta k/k in accordance with the Surveillance Frequency Control Program. This comparison shall consider at least those factors stated in Specification 4.1.1.1.1 e., above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPDs after each fuel loading. WATERFORD -UNIT 3 3/4 1-2 AMENDMENT NO ++,-249 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN -ALL CEAS FULLY INSERTED LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to that specified in the COLR. APPLICABILITY: MODES 3, 4 and 5 with all CEAs fully inserted. ACTION: With the SHUTDOWN MARGIN less than that specified in the COLR, immediately initiate boration to restore SHUTDOWN MARGIN to within limit. SURVEILLANCE REQUIREMENTS 4.1.1.2 With all CEAs fully inserted, the SHUTDOWN MARGIN shall be determined to be greater than or equal to that specified in the COLR, in accordance with the Surveillance Frequency Control Program by consideration of the following factors: 1. Reactor Coolant System boron concentration, 2. CEA position, 3. Reactor Coolant System average temperature, 4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration. WATERFORD -UNIT 3/4 1-3 AMENDMENT NO. 11.33,102, 141, REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tcoid) shall be greater than or equal to 533 °F. APPLICABILITY: MODES 1 and 2#. ACTION: With a Reactor Coolant System operating loop temperature (Tcoid) less than 533°F, restore Tco1d to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (T co1d) shall be determined to be greater than or equal to 533°F in accordance with the Surveillance Frequency Control Program. #\Nith Kett greater than or equal to 1.0. WATERFORD -UNIT 3 314 1-5 AMENDMENT NO. 2%, 249 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORA TION SYSTEMS FLOW PATHS-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source: a. A flow path from the boric acid makeup tank via either a boric acid makeup pump or a gravity feed connection and any charging pump to the Reactor Coolant System if the boric acid makeup tank in Specification 3.1.2. 7a. is OPERABLE, or b. The flow path from the refueling water storage pool via either a charging pump or a high pressure safety injection pump to the Reactor Coolant System if the refueling water storage pool in Specification 3.1.2.?b. is OPERABLE. APPLICABILITY: MODES 5 and 6. ACTION: With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.* SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN. WATERFORD -UNIT 3 3/4 1-6 AMENDMENT NO. 10, 185, 199, 249 REACTIVITY CONTROL SYSTEMS FLOW PATHS -OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two boron injection flow paths to the RCS via the charging pumps shall be OPERABLE. The following flow paths may be used: a. With the contents of either boric acid makeup tank in accordance with Figure 3.1-1, the following flow paths shall be OPERABLE: 1. One flow path from an acceptable boric acid makeup tank via its boric acid makeup pump; and 2. One flow path from an acceptable boric acid makeup tank via its gravity feed valve; or b. With the combined contents of both boric acid makeup tanks in dance with Figure 3.1-2, both of the following flow paths shall be OPERABLE: 1. One flow path consisting of both boric acid makeup pumps, and 2. One flow path consisting of both gravity feed valves. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to the requirements of Specification 3.1.1.1 or 3.1.1.2, whichever is applicable, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE: a. By verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position in accordance with the Surveillance Frequency Control Program. b. By verifying that each automatic valve in the flow path actuates to its correct position on an SIAS test signal in accordance with the Surveillance Frequency Control Program. c. By verifying that the flow path required by Specification 3.1.2.2a.1 and 3.1.2.2a.2 delivers at least 40 gpm to the Reactor Coolant System in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 1-7 AMENDMENT NO. 10, 199, 249 REACTIVITY CONTROL SYSTEMS CHARGING PUMPS -OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two independent charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to the requirements of Specification 3.1.1.1 or 3.1.1.2, whichever is applicable, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.1.2.4 Each required charging pump shall be demonstrated OPERABLE by verifying that each charging pump starts in response to an SIAS test signal in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 1-9 AMENDMENT NO . .i+, 249 REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS -OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid makeup pump(s) in the boron injection flow path(s) required OPERABLE pursuant to Specification 3.1.2.2a. shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump(s) in Specification 3.1.2.2a. is OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one boric acid makeup pump required for the boron injection flow path(s) pursuant to Specification 3.1.2.2a. inoperable, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to the ments of Specification 3.1.1.1 or 3.1.1.2, whichever is applicable, restore the above required boric acid makeup pump(s) to OPERABLE status within the next 7days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.1.2.6 Each required boric acid makeup pump shall be demonstrated OPERABLE by verifying that each boric acid makeup pump starts in response to an SIAS test signal in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 1-11 AMENDMENT NO. 44, 249 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE: a. One boric acid makeup tank with a boron concentration between 4900 ppm and 6125 ppm and a minimum borated water volume of 36% indicated level. b. The refueling water storage pool (RWSP) with: 1. A minimum contained borated water volume of 12% indicated level, and 2. A minimum boron concentration of 2050 ppm. APPLICABILITY: MODES 5 and 6. ACTION: With no borated water sources OPERABLE, suspend all operations involving CORE AL TERA TIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program when the Reactor Auxiliary Building air temperature is less than 55 ° F by verifying the boric acid makeup tank solution is greater than or equal to 60 ° F (when it is the source of borated water). b. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the boron concentration of the water, and 3. Verifying the contained borated water volume of the tank.

Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN. WATERFORD-UNIT 3 3/4 1-12 AMENDMENT NO. 10, 129, 185, 199, 249 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE: a. At least one of the following sources: 1) One boric acid makeup tank, with the tank contents in accordance with Figure 3.1-1, or 2) Two boric acid makeup tanks, with the combined contents of the tanks in accordance with Figure 3.1-2, and b. The refueling water storage pool in accordance with Specification 3.5.4. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. With the above required boric acid makeup tank(s) inoperable, restore the tank(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to the requirements of Specification 3.1.1.1 or 3.1.1.2, whichever is applicable; restore the above required boric acid makeup tank(s) to OPERABLE status within the next 7 days or be in . COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With the refueling water storage pool inoperable, restore the pool to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying the boric acid makeup tank solution temperature is greater than or equal to 60 ° F when the Reactor Auxiliary Building air temperature is less than 55 ° F. b. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the boron concentration in the water, and 2. Verifying the contained borated water volume of the water source. WATERFORD -UNIT 3 3/41-13 AMENDMENT NO. 10, 19, 129, 147, 199, 249 REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.1.2.9.1 The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 from MODE 2. 4.1.2.9.2 Each required boron dilution alarm shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK in accordance with the Surveillance Frequency Control Program, a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program, and a CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program. 4.1.2.9.3 If the primary makeup water flow path to the Reactor Coolant System is isolated to fulfill 3.1.2.9.b, the required primary makeup water flow path to the Reactor Coolant System shall be verified to be isolated by either locked closed manual valves, deactivated automatic valves secured in the isolation position, or by power being removed from all charging pumps, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 4.1.2.9.4 The requirements of Specification 3.1.2.9.a.2 or 3.1.2.9.b.2 shall be verified in accordance with the Surveillance Frequency Control Program. 4.1.2.9.5 Each required boron dilution alarm setpoint shall be adjusted to less than or equal to the existing neutron flux (cps) multiplied by the value specified in the COLR, at the frequencies specified in the COLR. WATERFORD -UNIT 3 3/4 1-16 AMENDMENT NO. Q, 48, 59, 102, 249 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each CEA shall be determined to be within 7 inches (indicated position) of all other CEAs in its group in accordance with the Surveillance Frequency Control Program except during time intervals when one CEAC is inoperable or when both CEACs are inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 4.1.3.1.2 Each CEA not fully inserted in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 314 1-20 AMENDEMENT NO. 87, 182, 249 REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA: a. CEA Reed Switch Position Transmitter (RSPT 1) with the capability of determining the absolute CEA positions within 5 inches, b. CEA Reed Switch Position Transmitter (RSPT 2) with the capability of determining the absolute CEA positions within 5 inches, and c. The CEA pulse counting position indicator channel. APPLICABILITY: MODES 1 and 2. ACTION: With a maximum of one CEA per CEA group having only one of the above required CEA position indicator channels OPERABLE, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either: a. Restore the inoperable position indicator channel to OPERABLE status, or b. Be in at least HOT STANDBY, or c. Position the CEA group(s) with the inoperable position indicator(s) at its fully withdrawn position while maintaining the requirements of Specifications 3.1.3.1 and 3.1.3.6. Operation may then continue provided the CEA group(s) with the inoperable position indicator(s) is maintained fully withdrawn, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2, and each CEA in the group(s) is verified fully withdrawn at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter by its "Full Out" limit. SURVEILLANCE REQUIREMENTS 4.1.3.2 Each of the above required position indicator channels shall be determined to be OPERABLE by verifying that for the same CEA, the position indicator channels agree within 5 inches of each other in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 1-21 AMENDMENT NO. 249 REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS -SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 At least one CEA Reed Switch Position Transmitter indicator channel shall be OPERABLE for each CEA not fully inserted. APPLICABILITY: MODES 3*, 4*, and 5*. ACTION: With less than the above required position indicator channel(s) OPERABLE, immediately open the reactor trip breakers. SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required CEA Reed Switch Position Transmitter indicator channel(s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable for performance of this surveillance testing. *With the reactor trip breakers in the closed position. WATERFORD -UNIT 3 3/4 1-22 AMENDMENT NO. 249 REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to greater than or equal to 145 inches. APPLICABILITY: MODES 1** and 2*#**. ACTION: With a maximum of one shutdown CEA withdrawn to less than 145 inches withdrawn, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either: a. Withdraw the CEA to greater than or equal to 145 inches, or b. Declare the CEA inoperable and determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to greater than or equal to 145 inches withdrawn: a. Within 15 minutes prior to withdrawal of any CEAs in regulating groups or group P during an approach to reactor criticality, and b. In accordance with the Surveillance Frequency Control Program. *See Special Test Exception 3.10.2. #With Keff greater than or equal to 1.0. **Except for surveillance testing pursuant to Specification 4.1.3.1.2. WATERFORD -UNIT 3 314 1-24 AMENDMENT NO. 249 REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION <Continued) ACTION: (Continued) c. With the regulating CEA groups or group P CEAs inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for intervals greater than 5 EFPD per 30 EFPD interval or greater than 14 EFPD per calendar year, either: 1. Restore the regulating CEA groups or group P CEAs to within the Long Term Steady State Insertion Limits within two hours, or 2. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group and CEA group P shall be determined to be within the Transient Insertion Limits in accordance with the Surveillance Frequency Control Program except during time intervals when the POil Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The accumulated times during which the regulating CEA groups or CEA group Pare inserted beyond the Long Term Steady State Insertion Limits but within the Transient Insertion Limits shall be determined in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 314 1-26 AMENDMENT NO. 249 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (COLSS) or, with the COLSS out of service, by verifying in accordance with the Surveillance Frequency Control Program that the linear heat rate, as indicated on any OPERABLE Local Power Density channel, is within the limits specified in the COLR. 4.2.1.3 In accordance with the Surveillance Frequency Control Program, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on kW/ft. WATERFORD -UNIT 3 314 2-1a AMENDMENT NO. 32, 102, 249 POWER DISTRIBUTION LIMITS 3/4.2.2 PLANAR RADIAL PEAKING FACTORS -LIMITING CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIAL PEAKING FACTORS (Fm xy) shall be less than or equal to the PLANAR RADIAL PEAKING FACTORS (Fcxy) used in the Core Operating Limit Supervisory System (COLSS) and in the Core Protection Calculators (CPC). APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.* ACTION: With a Fm xy exceeding a corresponding Fcxy. within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either: a. Adjust the CPC addressable constants to increase the multiplier applied to planar radial peaking by a factor equivalent to greater than or equal to Fm xy/Fc xy and restrict subsequent operation so that a margin to the COLSS operating limits of at least [Fm xy!Fcxy) -1.0) x 100% is maintained; or b. Adjust the affected PLANAR RADIAL PEAKING FACTORS (Fcxy) used in the COLSS and CPC to a value greater than or equal to the measured PLANAR RADIAL PEAKING FACTORS (Fm xy) or c. Be in at least HOT STANDBY. SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 The measured PLANAR RADIAL PEAKING FACTORS (Fm xy) obtained by using the incore detection system, shall be determined to be less than or equal to the PLANAR RADIAL PEAKING FACTORS (Fcxy). used in the COLSS and CPC at the following intervals: a. After each fuel loading with THERMAL POWER greater than 40% but prior to operation above 70% of RATED THERMAL POWER, and b. In accordance with the Surveillance Frequency Control Program. *See Special Test Exception 3.10.2. WATERFORD -UNIT 3 3/4 2-3 AMENDMENT NO. 249 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit above 20% of RA TED THERMAL POWER by: a. Continuously monitoring the tilt with COLSS when the COLSS is OPERABLE. b. Calculating the tilt in accordance with the Surveillance Frequency Control Program when the COLSS is inoperable. c. Verifying in accordance with the Surveillance Frequency Control Program, that the COLSS Azimuthal Tilt Alarm is actuated at an AZIMUTHAL POWER TILT greater than the AZIMUTHAL POWER TILT Allowance used in the CPCs. d. Using the incore detectors in accordance with the Surveillance Frequency Control Program to independently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT. WATERFORD -UNIT 3 3/4 2-5 AMENDMENT NO. 249 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable. 4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying in accordance with the Surveillance Frequency Control Program that the DNBR, as indicated on any OPERABLE DNBR channel, is within the limit specified in the COLR. 4.2.4.3 In accordance with the Surveillance Frequency Control Program, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR. WATERFORD -UNIT 3 3/4 2-6a AMENDMENT NO. 32, 102, 249 POWER DISTRIBUTION LIMITS 3/4.2.5 RCS FLOW RA TE LIMITING CONDITION FOR OPERATION 3.2.5 The actual Reactor Coolant System total flow rate shall be greater than or equal to 148.0 x 106 lbm/h. APPLICABILITY: MODE 1. ACTION: With the actual Reactor Coolant System total flow rate determined to be less than the above limit, reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.2.5 The actual Reactor Coolant System total flow rate shall be determined to be greater than or equal to the above limit in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 2-10 AMENDMENT NO. 249 POWER DISTRIBUTION LIMITS 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE LIMITING CONDITION FOR OPERATION 3.2.6 The reactor coolant cold leg temperature (Tc) shall be maintained between 536°F and 549°F.* APPLICABILITY: MODE 1 above 30% of RATED THERMAL POWER. ACTION: With the reactor coolant cold leg temperature exceeding its limit, restore the temperature to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 30% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.2.6 The reactor coolant cold leg temperature shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program. *Following a reactor power cutback in which (1) Regulating Groups 5 and/or 6 are dropped or (2) Regulating Groups 5 and/or 6 are dropped and the remaining Regulating Groups (Groups 1, 2, 3, and 4) are sequentially inserted, the upper limit on Tc may increase to 559°F for up to 30 minutes. WATERFORD -UNIT 3 3/4 2-11 AMENDMENT NO. 120, 199, 249 POWER DISTRIBUTION LIMITS 3/4.2.7 AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The AXIAL SHAPE INDEX (ASI) shall be maintained within the limits specified in the COLR. APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.* ACTION: With the AXIAL SHAPE INDEX outside the limits specified in the COLR, restore the AXIAL SHAPE INDEX to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 20% of RA TED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.2.7 The AXIAL SHAPE INDEX shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program using the COLSS or any OPERABLE Core Protection Calculator channel. *See Special Test Exception 3.10.2. WATERFORD -UNIT 3 3/4 2-12 AMENDMENT 13, 26, 102, 249 POWER DISTRIBUTION LIMITS 3/4.2.8 PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION 3.2.8 The steady-state pressurizer pressure shall be maintained between 2125 psia and 2275 psia. APPLICABILITY: MODE 1 ACTION: With the steady-state pressurizer pressure outside its above limits, restore the pressure to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.2.8 The steady-state pressurizer pressure shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 2-13 AMENDMENT NO. 4-99, 249 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1. 4.3.1.2 The logic for the bypasses shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total bypass function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit in accordance with the Surveillance Frequency Control Program. Neutron detectors are exempt from response time testing. Each test shall include at least one channel per function such that all channels are tested as shown in the "Total No. of Channels" column of Table 3.3-1. 4.3.1.4 The isolation characteristics of each CEA isolation amplifier and each optical isolator for CEA Calculator to Core Protection Calculator data transfer shall be verified in accordance with the Surveillance Frequency Control Program during the shutdown per the following tests: a. For the CEA position isolation amplifiers: 1. With 120 volts AC (60 Hz) applied for at least 30 seconds across the output, the reading on the input does not exceed 0.015 volts DC. WATERFORD -UNIT 3 3/4 3-1 AMENDMENT NO. 94, 249 INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) 2. With 120 volts AC (60 Hz) applied for at least 30 seconds across the input, the reading on the output does not exceed 15.0 volts DC. b. For the optical isolators: Verify that the input to output insulation resistance is greater than 10 megohms when tested using a megohmmeter on the 500 volt DC range. 4.3.1.5 The Core Protection Calculator System and the Control Element Assembly Calculator System shall be determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that less than three auto restarts have occurred on each calculator during the past 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 4.3.1.6 The Core Protection Calculator System shall be subjected to a CHANNEL FUNCTIONAL TEST to verify OPERABILITY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of receipt of a High CPC Cabinet Temperature alarm. WATERFORD -UNIT 3 3/4 3-2 AMENDMENT NO. 249 TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 1. Manual Reactor Trip N.A. N.A. SFCP and S/U(1) 1, 2, 3*, 4*, 5* 2. Linear Power Level -High SFCP SFCP(2,4),SFCP (3,4), SFCP 1, 2 SFCP (4) 3. Logarithmic Power Level -High SFCP SFCP(4) SFCP and S/U(1) . 2#, 3,4, 5 4. Pressurizer Pressure -High SFCP SFCP SFCP 1, 2 5. Pressurizer Pressure -Low SFCP SFCP SFCP 1, 2 6. Containment Pressure -High SFCP SFCP SFCP 1, 2 7. Steam Generator Pressure -Low SFCP SFCP SFCP 1, 2 8. Steam Generator Level -Low SFCP SFCP SFCP 1, 2 9. Local Power Density -High SFCP SFCP(2,4 ),SFCP(4,5) SFCP, SFCP(6) 1, 2 10. DNBR-Low SFCP SFCP(7), SFCP(2,4), SFCP, SFCP(6) 1, 2 SFCP(8), SFCP(4,5) 11. DELETED 12. Reactor Protection System Logic N.A. N.A. SFCP(11) and S/U(1) 1, 2, 3*, 4*, 5* WATERFORD -UNIT 3 314 3-10 AMENDMENT NO. 40. 69. 153. Ž* 249 TABLE 4.3-1 (Continued) REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 13. Reactor Trip Breakers N.A. N.A. SFCP(10, 11 ), S/U(1) 1, 2, 3*, 4*, 5* 14. Core Protection Calculators SFCP SFCP(2,4), SFCP(9), SFCP(6) 1,2 SFCP(4,5) 15. CEA Calculators SFCP SFCP SFCP, SFCP(6) 1, 2 16. Reactor Coolant Flow -Low SFCP SFCP SFCP 1, 2 WATERFORD -UNIT 3 3/4 3-11 AMENDMENT NO. 69. 153, 249 TABLE 4.3-1 (Continued) TABLE NOTATIONS (Continued) (3) Above 15% of RATED THERMAL POWER, verify that the linear power subchannel gains of the excore detectors are consistent with the values used to establish the shape annealing matrix elements in the Core Protection Calculators. (4) Neutron detectors may be excluded from CHANNEL CALIBRATION. (5) After each fuel loading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be used to determine or verify acceptable values for the shape annealing matrix elements used in the Core Protection Calculators. (6) This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to sensors as practicable to verify OPERABILITY including alarm and/or trip functions. (7) Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation or by calorimetric calculations and if necessary, adjust the CPC addressable constant flow co-efficients such that each CPC indicated flow is less than or equal to the actual flow rate. The flow measurement uncertainty is included in the BERR1 term in the CPC and is equal to or greater than 4%. (8) Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by calorimetric calculations. (9) The CHANNEL FUNCTIONAL TEST shall include verification that the correct values of addressable constants are installed in each OPERABLE CPC. (10) In accordance with the Surveillance Frequency Control Program and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage trip function and the shunt trip function. (11) The CHANNEL FUNCTIONAL TEST shall be scheduled and performed such that the Reactor Trip Breakers (RTBs) are tested at least every 6 weeks to accommodate the appropriate vendor recommended interval for cycling of each RTB. WATERFORD -UNIT 3 3/4 3-12a AMENDMENT NO. 125,153, 222, 249 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. APPLICABILITY: As shown in Table 3.3-3. ACTION: a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value. b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3. SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2. 4.3.2.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit in accordance with the Surveillance Frequency Control Program. Each test shall include at least one channel per function such that all channels are tested as shown in the "Total No. of Channels" Column of Table 3.3-3. WATERFORD -UNIT 3 3/4 3-13 AMENDMENT NO. 94, 249 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 1. SAFETY INJECTION (SIAS) a. Manual (Trip Buttons) N.A. N.A. SFCP 1, 2, 3,4 b. Containment Pressure -High SFCP SFCP SFCP 1, 2, 3 c. Pressurizer Pressure -Low SFCP SFCP SFCP 1, 2, 3 d. Automatice Actuation Logic (except subgroup relays) N.A. N.A. SFCP(2) 1, 2, 3 Actuation Subgroup Relays N.A. N.A. SFCP(3) (6) 1, 2, 3 2. CONTAINMENT SPRAY (CSAS) a. Manual (Trip Buttons) N.A. N.A. SFCP 1, 2, 3,4 b. Containment Pressure --High -High SFCP SFCP SFCP 1, 2, 3 c. Automatic Actuation Logic (except subgroup relays) N.A. N.A. SFCP(2) 1, 2, 3 Actuation Subgroup Relays N.A. N.A. SFCP(1) (3) 1, 2, 3 3. CONTAINMENT ISOLATION (CIAS) a. Manual CIAS (Trip Buttons) N.A. N.A. SFCP 1, 2, 3,4 b. Containment Pressure -High SFCP SFCP SFCP 1, 2, 3 c. Pressurizer Pressure -Low SFCP SFCP SFCP 1, 2, 3 d. Automatic Actuation Logic (except subgroup relays) N.A. N.A. SFCP(2) 1, 2, 3 Actuation Subgroup Relays N.A. N.A. SFCP(1) (3) 1, 2, 3 4. MAIN STEAM LINE ISOLATION a. Manual (Trip Buttons) N.A. N.A. SFCP 1, 2, 3 b. Steam Generator Pressure -Low SFCP SFCP SFCP 1, 2, 3 c. Containment Pressure -High SFCP SFCP SFCP 1, 2, 3 d. Automatic Actuation Logic (except subgroup relays) N.A. N.A. SFCP(2) 1, 2, 3 Actuation Subgroup Relays N.A. N.A. SFCP(1) (3) 1, 2, 3 WATERFORD -UNIT 3 3/4 3-25 AMENDMENT NO. 67. 69. 78, 249 TABLE 4.3-2 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 5. SAFETY INJECTION SYSTEM RECIRCULATION (RAS) a. Manual RAS (Trip Buttons) N.A. N.A. SFCP 1, 2,3,4 b. Refueling Water Storage Pool-Low SFCP SFCP SFCP 1,2,3,4 c. Automatic Actuation Logic (except subgroup relays) N.A. N.A. SFCP(2) 1,2,3,4 Actuation Subgroup Relays N.A. N.A. SFCP(1) (3) 1,2,3,4 6. LOSS OF POWER (LOV) a. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) N.A. SFCP SFCP(4) 1, 2, 3 b. 480 V Emergency Bus Undervoltage (Loss of Voltage) N.A. SFCP SFCP(4) 1, 2, 3 c. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) N.A. SFCP SFCP(4) 1, 2, 3 WATERFORD -UNIT 3 3/4 3-26 AMENDMENT NO. 69. 78. 136, 249 TABLE 4.3.-2 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED 7. EMERGENCY FEEDWATER (EFAS) a. Manual (Trip Buttons) N.A. N.A. SFCP 1, 2, 3 b. SG Level (1/2) -Low and t;P (1/2) -High SFCP SFCP SFCP 1, 2, 3 c. SG Level (1/2) -Low and No Pressure -Low Trip (1/2) SFCP SFCP SFCP 1, 2, 3 d. Automatic Actuation Logic (except subgroup relays) N.A. N.A. SFCP(2) 1,2,3 Actuation Subgroup Relays N.A. N.A SFCP(1) (3) 1,2,3 e. Control Valve Logic SFCP SFCP SFCP(5) 1,2,3 (Wide Range SG Level -Low) TABLE NOTATION (1) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program. (2) Testing of Automatic Actuation Logic shall include the energization/deenergization of each initiation relay and verification of the OPERABILITY of each initiation relay. (3) A subgroup relay test shall be performed which shall include the energization/deenergization of each subgroup relay and verification of the OPERABILITY of each subgroup relay. Relays K109, K114, K202, K301, K305, K308 and K313 are exempt from testing during power operation but shall be tested in accordance with the Surveillance Frequency Control Program and during each COLD SHUTDOWN condition unless tested within the previous 62 days (4) Using installed test switches. (5) To be performed during each COLD SHUTDOWN if not performed in the previous 6 months. (6) Each train shall be tested, with the exemption of relays, K110, K410 and K412, in accordance with the Surveillance Frequency Control Program. Relays K110, K410 and K412 shall be tested in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 3-27 AMENDMENT NO 67. 69. 78, 249 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT 1. AREA MONITORS a. Deleted b. Containment -Purge & Exhaust Isolation 2. PROCESS MONITORS a. DELETED b. Control Room Intake Monitors c. Steam Generator Slowdown d. Component Cooling Water Monitors A&B e. Component Cooling Water Monitor A/B *Deleted CHANNEL CHECK SFCP SFCP SFCP SFCP SFCP CHANNEL CALIBRATION SFCP SFCP SFCP SFCP SFCP CHANNEL FUNCTIONAL TEST SFCP SFCP SFCP SFCP SFCP **During CORE AL TERA TIONS or load movements with or over irradiated fuel within the containment. ***During load movements with or over irradiated fuel. MODES FOR WHICH SURVEILLANCE IS REQUIRED 1, 2, 3, 4 & ** ALL MODES & *** 1, 2, 3, & 4 ALL MODES 1, 2, 3, & 4 WATERFORD -UNIT 3 3/4 3-32 AMENDMENT NO. 91. 96. 149.176. 197. 2aa, 249 TABLE 4.3-3 (Continued) RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST IS REQUIRED 3. EFFLUENT ACCIDENT MONITORS a. Containment High Range SFCP SFCP SFCP 1, 2, 3, & 4 b. Plant Stack High Range SFCP SFCP SFCP 1, 2, 3, & 4 c. Condenser Vacuum Pump High Range SFCP SFCP SFCP 1, 2, 3, & 4 d. Fuel Handling Building Exhaust High Range SFCP SFCP SFCP 1*, 2*, 3*, & 4* e. Main Steam Line High Range SFCP SFCP SFCP 1, 2, 3, & 4 *With irradiated fuel in the storage pool. WATERFORD -UNIT 3 3/4 3-33 AMENDMENT NO. 99, 249 TABLE 4.3-6 REMOTE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQURIEMENTS CHANNEL CHANNEL INSTRUMENTATION CHECK CALIBRATION 1. Neutron Flux SFCP SFCP* 2. Reactor Trip Breaker Indication SFCP N.A. 3. Reactor Coolant Temperature -Cold Leg (T Cold) SFCP SFCP 4. Reactor Coolant Temperature -Hot Leg (T Hot) SFCP SFCP 5. Pressurizer Pressure SFCP SFCP 6. Pressurizer Level SFCP SFCP 7. Steam Generator Level SFCP SFCP 8. Steam Generator Pressure SFCP SFCP 9. Shutdown Cooling Flow Rate SFCP SFCP 10. Emergency Feedwater Flow Rate SFCP SFCP 11. Condensate Storage Pool Level SFCP SFCP *Neutron detector may be excluded from CHANNEL CALIBRATION. WATERFORD -UNIT 3 314 3-43 AMENDMENT NO. 249 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION 1. Containment Pressure (Wide Range) SFCP SFCP 2. Containment Pressure (Wide Wide Range) SFCP SFCP 3. Reactor Coolant Outlet Temperature -T Hot (Wide Range) SFCP SFCP 4. Reactor Coolant Inlet Temperature -T co1d (Wide Range) SFCP SFCP 5. Reactor Coolant Pressure -Wide Range SFCP SFCP 6. Pressurizer Water Level SFCP SFCP 7. Steam Generator Water Level -Narrow Range SFCP SFCP 8. Steam Generator Water Level -Wide Range SFCP SFCP 9. Containment Water Level (Wide Range) SFCP SFCP 10. Core Exit Thermocouples SFCP SFCP 11. Containment Isolation Valve Position SFCP SFCP 12. Condensate Storage Pool Level SFCP SFCP 13. Reactor Vessel Level Monitoring System SFCP SFCP 14. Log Power Indication (Neutron Flux) SFCP SFCP WATERFORD -UNIT 3 3/4 3-46 AMENDMENT NO. 44 +n,. 249 INSTRUMENTATION CHEMICAL DETECTION SYSTEMS CHLORINE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.7.1 Two independent chlorine detection systems, with their alarm/trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 2 ppm, shall be OPERABLE. APPLICABILITY: All MODES. ACTION: a. With one chlorine detection system inoperable, restore the inoperable detection system to OPERABLE status within 7 days or within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months initiate and maintain operation of the control room ventilation system in the isolate mode of operation. b. With no chlorine detection system OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the control room ventilation system in the isolate mode of operation. c. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.7.1 Each chlorine detection system shall be demonstrated OPERABLE by performance of a CHANNEL CHECK in accordance with the Surveillance Frequency Control Program and a CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 3-47 AMENDMENT NO. 21, 53, 156, 249 Correstion Letter of a 17 2000 INSTRUMENTATION CHEMICAL DETECTION SYSTEMS BROAD RANGE GAS DETECTION LIMITING CONDITION FOR OPERATION 3.3.3.7.3 Two independent broad range gas detection systems shall be OPERABLE** with their alarm/trip setpoints adjusted to actuate at the lowest achievable Immediately Dangerous to Life or Health gas concentration level of detectable toxic gases* providing reliable operation. APPLICABILITY: All MODES. ACTION: a. With one broad range gas detection system inoperable, restore the inoperable detection system to OPERABLE status within 7 days or within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months initiate and maintain operation of the control room ventilation system in the isolate mode of operation. b. With no broad range gas detection system OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the control room ventilation system in the isolate mode of operation. c. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.7.3 Each broad range gas detection system shall be demonstrated OPERABLE by performance of a CHANNEL CHECK in accordance with the Surveillance Frequency Control Program, and a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program. The CHANNEL FUNCTIONAL TEST will include the introduction of a standard gas. *Including Ammonia ** The requirements of Technical Specification 3.0.1 do not apply during the time (two minutes or less) when the instrument automatic background/reference spectrum check renders the instrument(s) inoperable. WATERFORD -UNIT 3 3/4 3-48a AMENDMENT NO. 20,53,133,135,151, 249 TABLE 4.3-9 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED 1. WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM a. Hydrogen Monitor SFCP N.A. SFCP(4) SFCP ** b. Oxygen Monitors SFCP N.A. SFCP(5) SFCP ** WATERFORD -UNIT 3 3/4 3-65 AMENDMENT NO. eg, 249 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation. APPLICABILITY: MODES 1 and 2. ACTION: With less than the above required reactor coolant pumps in operation, be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 4-1 AMENDMENT NO. 249 REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 The reactor coolant loops listed below shall be OPERABLE and at least one of these reactor coolant Loops shall be in operation.* a. Reactor Coolant Loop 1 and its associated steam generator and at least one associated reactor coolant pump. b. Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump. APPLICABILITY: MODE 3**. ACTION: a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . b. With no reactor coolant loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2 and immediately initiate corrective action to return the required reactor coolant loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 At least one reactor coolant loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. 4.4.1.2.3 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be z 50% of wide range indication in accordance with the Surveillance Frequency Control Program. *All reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2, and (2) core outlet temperature is maintained at least 10°F below saturation temperature. **See Special Test Exception 3.10.5. WATERFORD -UNIT 3 3/4 4-2 AMENDMENT NO. 249 REACTOR COOLANT SYSTEM HOT SHUTDOWN SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability. 4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be ::> 50% of wide range indication in accordance with the Surveillance Frequency Control Program. 4.4.1.3.3 At least one reactor coolant or shutdown cooling loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 4-4 AMENDMENT NO. 249 REACTOR COOLANT SYSTEM COLD SHUTDOWN -LOOPS FILLED SURVEILLANCE REQUIREMENTS 4.4.1.4.1 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability. 4.4.1.4.2 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be ;;.:50% of wide range indication in accordance with the Surveillance Frequency Control Program 4.4.1.4.3 At least one reactor coolant loop or shutdown cooling train shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 4-5a AMENDMENT N0.- 96, 249 Cor.r:eo#oR letter of 7 23 2()()4 REACTOR COOLANT SYSTEM COLD SHUTDOWN -LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.5 Two shutdown cooling loops shall be OPERABLE# and at least one shutdown cooling loop shall be in operation.* APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION: a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. b. With no shutdown cooling loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2 and immediately initiate corrective action to return the required shutdown cooling loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.5 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. # One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other shutdown cooling loop is OPERABLE and in operation. *The shutdown cooling pump (LPSI pump) may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.1 or 3.1.1.2, and (2) core outlet temperature is maintained at least 10°F below saturation temperature. WATERFORD -UNIT 3 3/4 4-6 AMENDMENT NO. 249 REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3.1 The pressurizer shall be OPERABLE with: a. A steady-state water volume greater than or equal to 26% indicated level (350 cubic feet) but less than or equal to 62.5% indicated level (900 cubic feet), and, b. At least two groups of pressurizer heaters powered from Class 1 E buses each having a nominal capacity of 150 kW. APPLICABILITY: MODES 1, 2, and 3. ACTION: a. With only one group of the above required pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.4.3.1.1 The pressurizer water volume shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program. 4.4.3.1.2 The capacity of each of the above required groups of pressurizer heaters shall be verified to be at least 150 kW in accordance with the Surveillance Frequency Control Program. 4.4.3.1.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by: a. Verifying the above pressurizer heaters are automatically shed from the emergency power sources upon the injection of an SIAS test signal. b. Verifying that the above heaters can be manually placed and energized on the emergency power source from the control room. WATERFORD -UNIT 3 3/4 4-9 Amendment No. 249 REACTOR COOLANT SYSTEM AUXILIARY SPRAY LIMITING CONDITION FOR OPERATION 3.4.3.2 Both auxiliary spray valves shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: a. With only one of the above required auxiliary spray valves OPERABLE, restore both valves to OPERABLE status within 30 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . b. With none of the above required auxiliary spray valves OPERABLE, restore at least one valve to OPERABLE status within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The auxiliary spray valve shall be verified to have power available to each valve in accordance with the Surveillance Frequency Control Program. 4.4.3.2.2 The auxiliary spray valves shall be cycled in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 4-9a Amendment No. 2 , 249 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION <Continued) Perform SR 4.4.5.2.1 once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and restore the containment sump monitor to OPERABLE status within 30 days; Be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . c. All required RCS leakage detection instrumentation inoperable. Initiate ACTION within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.4.5.1 The leakage detection systems shall be demonstrated OPERABLE by: a. Containment atmosphere particulate monitor system -performance of CHANNEL CHECK in accordance with the Surveillance Frequency Control Program, CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program and CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program. b. Containment sump level and flow monitors -performance of a CHANNEL CHECK (containment sump level monitor only) in accordance with the Surveillance Frequency Control Program and a CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 4-17a AMENDMENT NO. 197,212, 249 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.5.2 Reactor Coolant System operational leakage shall be limited to: a. No PRESSURE BOUNDARY LEAKAGE, b. 1 gpm UNIDENTIFIED LEAKAGE, c. 75 gallons per day primary to secondary leakage, through any one steam generator (SG), d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and e. 1 gpm leakage at a Reactor Coolant System pressure of 2250 +/- 20 psia from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. With any PRESSURE BOUNDARY LEAKAGE, or primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With any Reactor Coolant System operational leakage greater than any one of the limits, excluding PRESSURE BOUNDARY LEAKAGE, primary to secondary leakage, and leakage from Reactor Coolant System pressure isolation valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . c. With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual or deactivated automatic valve, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS NOTE: Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. 4.4.5.2.1 Reactor Coolant System leakages, except for primary to secondary leakage, shall be demonstrated to be within each of the above limits by performance of a Reactor Coolant System water inventory balance in accordance with the Surveillance Frequency Control Program. 4.4.5.2.2 Primary to secondary leakage shall be verified to be gallons per day through any one SG in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 4-18 AMENDMENT NO. 197, 199, 204, 249 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.2.3 Each Reactor Coolant System pressure isolation valve specified in Table 3.4-1, Section A and Section B, shall be demonstrated OPERABLE by verifying leakage to be within its limit: a. In accordance with the Surveillance Frequency Control Program, b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, c. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, d. Following valve actuation for valves in Section B due to automatic or manual action or flow through the valve: 1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying valve closure, and 2. Within 31 days by verifying leakage rate. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. 4.4.5.2.4 Each Reactor Coolant System pressure isolation valve power-operated valve specified in Table 3.4-1, Section C, shall be demonstrated OPERABLE by verifying leakage to be within its limit: a. In accordance with the Surveillance Frequency Control Program, and b. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. WATERFORD-UNIT 3 3/4 4-19 AMENDMENT NO 96, 107, 204, 249 TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT AND ANALYSIS 1 . Gross Activity Determination 2. Isotopic Analysis for DOSE EQUIVALENT 1-131 Concentration 3. Radiochemical for E Determination 4. Isotopic Analysis for Iodine Including 1-131, 1-133, and 1-135 SAMPLE AND ANALYSIS FREQUENCY a) b) SFCP SFCP SFCP* Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds 1.0 µCi/gram, DOSE EQUIVALENT 1-131 or 100/E µCi/gram, and One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15 % of the RATED THERMAL POWER within a 1-hour period. MODES IN WHICH SAMPLE AND ANALYSIS REQUIRED 1, 2, 3, 4 1 1 1#,2#,3#,4#, 5# 1, 2, 3 *Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer. # Until the specific activity of the primary coolant system is restored within its limits. WATERFORD -UNIT 3 3/4 4-25 AMENDMENT NO. +84, 249 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR Part 50 Appendix H in accordance with the Reactor Vessel material surveillance program -withdrawal schedule in FSAR Table 5.3-10. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. WATERFORD -UNIT 3 314 4-29 AMENDMENT NO. 106, 177, 249 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8.3.1 For each SOC System suction line relief valve: a. verify in the control room in accordance with the Surveillance Frequency Control Program that each valve in the suction path between the RCS and the SOC relief valve is open. b. verify each SOC relief valve is OPERABLE in accordance with the lnservice Testing Program. 4.4.8.3.2 With the RCS vented per ACTIONS a, b, or c, the RCS vent(s) and all valves in the vent path shall be verified to be open in accordance with the Surveillance Frequency Control Program*. *Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 4-35 AMENDMENT NO. 66, 72, 140, 189, 249 REACTOR COOLANT SYSTEM 3/4.4.10 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3 .4.10 At least one Reactor Coolant System vent path consisting of at least two valves in series powered from emergency buses shall be OPERABLE and closed at each of the following locations: a. Reactor vessel head, and b. Pressurizer steam space. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves and block valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With two or more Reactor Coolant System vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the vent valves and block valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.4.10 Each Reactor Coolant System vent path shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by: a. Verifying all manual isolation valves in each vent path are locked in the open position. b. Cycling each vent valve through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING. c. Verifying flow through the Reactor Coolant System vent paths during venting during COLD SHUTDOWN or REFUELING. WATERFORD -UNIT 3 3/4 4-37 AMENDMENT NO. 249 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ACTION: (Continued) MODES 1, 2, 3 and 4 with pressurizer pressure greater than or equal to 1750 psia (continued). d. With two of the required safety injection tanks inoperable, restore one of the tanks to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1750 psia within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . MODES 3 and 4 with pressurizer pressure less than 1750 psia e. With one of the required safety injection tanks inoperable due to boron concentration not within limits, restore the boron concentration to within limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . f. With one of the required safety injection tanks inoperable due to inability to verify level or pressure, restore the tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . g. With one of the required safety injection tanks inoperable for reasons other than ACTION a orb, restore the inoperable tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or be in at least COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . h. With two of the required safety injection tanks inoperable, restore one of the tanks to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or be in at least COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . SURVEILLANCE REQUIREMENTS 4.5.1 Each safety injection tank shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and 2. Verifying that each safety injection tank isolation valve is open. b. In accordance with the Surveillance Frequency Control Program by verifying the boron concentration of the safety injection tank solution. c. Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the safety injection tank solution. This surveillance is not required when the volume increase makeup source is the RWSP. WATERFORD -UNIT 3 3/4 5-2 AMENDMENT NO. 4-§e, 249 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) SURVEILLANCE REQUIREMENTS (Continued) d. In accordance with the Surveillance Frequency Control Program when the RCS pressure is above 1750 psia, by verifying that the isolation valve operator breakers are padlocked in the open position. e. In accordance with the Surveillance Frequency Control Program by verifying that each safety injection tank isolation valve opens automatically under each of the following conditions: 1. When an actual or simulated RCS pressure signal exceeds 535 psia, and 2. Upon receipt of a safety injection test signal. WATERFORD -UNIT 3 314 5-2a AMENDMENT NO. 249 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying that the following valves are in the indicated positions with the valves key-locked shut: Valve Number Valve Functions Valve Position a. 2Sl-V1556 a. Hot Leg Injection a. SHUT (Sl-506A) b. 2Sl-V1557 b. Hot Leg Injection b. SHUT (Sl-502A) c. 2Sl-V1558 C. Hot Leg Injection c. SHUT (Sl-502B) d. 2Sl-V1559 d. Hot Leg Injection d. SHUT (Sl-506B) b. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. 2. Verifying the ECCS piping is full of water. c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the safety injection system sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed: 1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2. Of the areas affected within containment at the completion of containment entry when CONTAINMENT INTEGRITY is established. d. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the action of the open permissive interlock (OPI) and isolation valve position alarms of the shutdown cooling system when the reactor coolant system pressure (actual or simulated) is between 392 psia and 422 psia. WATERFORD -UNIT 3 3/4 5-4 AMENDMENT NO. 65, 130, 249 GorroetioR Lotter of 7 19 Q7 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2. A visual inspection of the safety injection system sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion. 3. Verifying that a minimum total of 380 cubic feet of granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets. 4. Verifying that when a representative sample of 13.07 +/- 0.03 grams of TSP from a TSP storage basket is submerged, without agitation, in 4 +/- 0.1 liters of 120 +/- 10°F water borated to 3011 +/- 30 ppm, the pH of the mixed solution is raised to greater than or equal to 7 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . e. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that each automatic valve in the flow path actuates to its correct position on SIAS and RAS test signals. 2. Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal: a. High pressure safety injection pump. b. Low pressure safety injection pump. 3. Verifying that on a recirculation actuation test signal, the low pressure safety injection pumps stop, the safety injection system sump isolation valves open. f. By verifying that each of the following pumps required to be OPERABLE performs as indicated on recirculation flow when tested pursuant to the lnservice Testing Program: 1. High pressure safety injection pump differential pressure greater than or equal to 1429 psid. 2. Low pressure safety injection pump differential pressure greater than or equal to 168 psid. WATERFORD -UNIT 3 3/4 5-5 AMENDMENT NO. 64, 127, 162,189, 209, 249 EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE POOL LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage pool shall be OPERABLE with: a. A minimum contained borated water volume of 83% indicated level, b. Between 2050 and 2900 ppm of boron, and c. A solution temperature of greater than or equal to 55°F and less than or equal to 100°F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the refueling water storage pool inoperable, restore the pool to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.5.4 The RWSP shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the contained borated water volume in the pool, and 2. Verifying the boron concentration of the water. b. In accordance with the Surveillance Frequency Control Program by verifying the RWSP temperature when the RAB air temperature is less than 55°F or greater than 100°F. WATERFORD -UNIT 3 3/4 5-9 AMENDMENT NO. 19, 129, 147, 199, 249 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated: a. In accordance with the Surveillance Frequency Control Program by verifying that all penetrations* not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except for valves that are open under administrative control as permitted by Specification 3.6.3. b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3. c. After each closing of each penetration subject to Type B testing, except containment air locks, if opened following a Type A or B test, by leak rate testing the seal in accordance with the Containment Leakage Rate Testing Program. *Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. WATERFORD -UNIT 3 3/4 6-1 AMENDMENT NO. +-&,-+24, 249 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE: a. By verifying seal leakage in accordance with the Containment Leakage Rate Testing Program, b. By conducting overall air lock leakage tests in accordance with the Containment Leakage Rate Testing Program. c. In accordance with the Surveillance Frequency Control Program by verifying that only one door in each air lock can be opened at a time. WATERFORD -UNIT 3 3/4 6-10 AMENDMENT NO. 249 CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained less than 27 inches H 20 gauge and greater than 14.275 psia. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 6-11 Amendment No. 'J+, 4-74, 249 CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall 95°F* and.::, 120 °F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. If the minimum containment average air temperature is less than 95°F* but greater than or equal to 90°F*, then within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either restore containment air temperature to greater than or equal to 95°F or reduce the peak linear heat generation rate limit in accordance with the COLR. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. If the minimum containment average air temperature is less than 90°F, then restore containment air temperature to greater than or equal to 95°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . c. If maximum containment average air temperature is greater than 120°F, then restore containment air temperature to less than or equal to 120°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at any three of the following locations and shall be determined in accordance with the Surveillance Frequency Control Program: Location a. Containment Fan Cooler No. 1A Air Intake b. Containment Fan Cooler No. 1 B Air Intake c. Containment Fan Cooler No. 1 C Air Intake d. Containment Fan Cooler No. 1 D Air Intake

The minimum containment average air temperature limit is only applicable at greater than 70% RATED THERMAL POWER. WATERFORD -UNIT 3 3/4 6-13 AMENDMENT NO. 199, 214, 249 Go::.ceeteEI Bf ElatoEI Ma;' 9, CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1. 7 Each containment purge supply and exhaust isolation valve (CAP 103, CAP 104, CAP 203, and CAP 204) shall be OPERABLE and may be open at no greater than the 52° open position allowed by the mechanical stop for less than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. With a containment purge supply and/or exhaust isolation valve(s) open for greater than or equal to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days at any open position, close the open valve(s) or isolate the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With a containment purge supply and/or exhaust isolation valve(s) having a measured leakage rate exceeding the limits of Surveillance Requirement 4.6.1.7.2, restore the inoperable valve(s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.1.7.1 The cumulative time that the purge supply or exhaust isolation valves are open during the past 365 days shall be determined in accordance with the Surveillance Frequency Control Program. 4.6.1.7.2 Each containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE in accordance with the Containment Leakage Rate Testing Program. 4.6.1.7.3 Each containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that the mechanical stops limit the valve opening to a position < 52° open. WATERFORD -UNIT 3 314 6-15 Amendment No. 124, 213, 249 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZA TION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWSP on a containment spray actuation signal and automatically transferring suction to the safety injection system sump on a recirculation actuation signal. Each spray system flow path from the safety injection system sump shall be via an OPERABLE shutdown cooling heat exchanger. APPLICABILITY: MODES 1, 2, 3, and 4*. ACTION: a. With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With two containment spray systems inoperable, restore at least one spray system to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying that the water level in the containment spray header riser is> 149.5 feet MSL elevation. b. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is correctly positioned to take suction from the RWSP. c. By verifying, that on recirculation flow, each pump develops a total head of greater than or equal to 219 psid when tested pursuant to the lnservice Testing Program. *With Reactor Coolant System pressure > 400 psia. WATERFORD -UNIT 3 314 6-16 AMENDMENT NO. 89, 163, 89, 249 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS (Continued) SURVEILLANCE REQUIREMENTS (Continued) d. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that each automatic valve in the flow path actuates to its correct position on a CSAS test signal. 2 Verifying that upon a recirculation actuation test signal, the safety injection system sump isolation valves open and that a recirculation mode flow path via an OPERABLE shutdown cooling heat exchanger is established. 3. Verifying that each spray pump starts automatically on a CSAS test signal. e. In accordance with the Surveillance Frequency Control Program by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed. WATERFORD -UNIT 3 3/4 6-17 AMENDMENT NO. 89, 163, 200, 249 CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 Two independent trains of containment cooling shall be OPERABLE with one fan cooler to each train. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one train of containment cooling inoperable, restore the inoperable train to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable containment cooling train to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.2.2 Each train of containment cooling shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by: 1. Starting each operational fan not already running from the control room and verifying that each operational fan operates for at least 15 minutes. 2. Verifying a cooling water flow rate of greater than or equal to 625 gpm to each cooler. b. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that each fan starts automatically on an SIAS test signal. 2. Verifying a cooling water flow rate of greater than or equal to 1200 gpm to each cooler. 3. Verifying that each cooling water control valve actuates to its full open position on a SIAS test signal. WATERFORD -UNIT 3 314 6-18 Amendment No. 39, 131, 165, 249 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) 4.6.3.2 Each containment isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by: a. Verifying that on a containment isolation test signal, each isolation valve actuates to its isolation position. b. Verifying that on a containment Radiation-High test signal, each containment purge valve actuates to its isolation position. 4.6.3.3 The isolation time of each power-operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to the lnservice Testing Program. WATERFORD -UNIT 3 3/4 6-20 AMENDMENT NO. 75, 189, 249 CONTAINMENT SYSTEMS 3/4.6.6 SECONDARY CONTAINMENT SHIELD BUILDING VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.1 Two independent shield building ventilation systems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one shield building ventilation system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.6.1 Each shield building ventilation system shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months continuous with the heaters on. b. In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by: WATERFORD-UNIT 3 3/4 6-37 AMENDMENT NO. 249 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1. Verifying that the ventilation system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 10,000 cfm +/- 10%. 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 70%. 3. Verifying a system flow rate of 10,000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975. c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 70%. d. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7 .8 inches water gauge while operating the system at a flow rate of 10,000 cfm +/- 10%. 2. Verifying that the system starts on a safety injection actuation test signal. 3. Verifying that the filter cooling bypass valves can be manually cycled. 4. Verifying that each system produces a negative pressure of greater than or equal to 0.25 inch water gauge in the annulus within 1 minute after a start signal. 5. Verifying that the heaters dissipate 60 + 6.0, -6.0 kW when tested in accordance with ANSI N510-1975. WATERFORD -UNIT 3 314 6-38 AMENDMENT NO. -i94, 249 CONTAINMENT SYSTEMS SHIELD BUILDING INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.2 SHIELD BUILDING INTEGRITY shall be maintained with an annulus negative pressure greater than 5 inches water gauge. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: Without SHIELD BUILDING INTEGRITY, restore SHIELD BUILDING INTEGRITY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.6.6.2 SHIELD BUILDING INTEGRITY shall be demonstrated: a. In accordance with the Surveillance Frequency Control Program by verifying the annulus pressure to be within its limits. b. In accordance with the Surveillance Frequency Control Program by verifying that each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, then at least one door shall be closed. WATERFORD -UNIT 3 3/4 6-40 AMENDMENT NO. 249 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.1.2 The emergency feedwater system shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying that each manual, power-operated, and automatic valve in each water flow path and in both steam supply flow paths to the turbine-driven EFW pump steam turbine, that is not locked, sealed, or otherwise secured in position, is in its correct position. b. At least once per 92 days by testing the EFW pumps pursuant to the lnservice Testing Program. This surveillance requirement is not required to be performed for the turbine-driven EFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 750 psig in the steam generators. c. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an actual or simulated actuation signal. NOTE: This surveillance requirement is not required to be performed for the turbine-driven EFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 750 psig in the steam generators. 2. Verifying that each EFW pump starts automatically upon receipt of an actual or simulated actuation signal. d. Prior to entering MODE 2, whenever the plant has been in MODE 4, 5, 6 or defueled, for 30 days or longer, or whenever feedwater line cleaning through the emergency feedwater line has been performed, by verifying flow from the condensate storage pool through both parallel flow legs to each steam generator. WATERFORD -UNIT 3 3/4 7-5 AMENDMENT NO. 96, 173, 189, 249 PLANT SYSTEMS CONDENSATE STORAGE POOL LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage pool (CSP) shall be OPERABLE with: a. A minimum contained volume of at least 92% indicated level.* b. A water temperature of greater than or equal to 55°F and less than or equal to 100°F. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: In MODES 1. 2. and 3: With the condensate storage pool inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the CSP to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In MODE 4: With the condensate storage pool inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the CSP to OPERABLE status or be in at least COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . SURVEILLANCE REQUIREMENTS 4. 7.1.3.1 The condensate storage pool shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying the contained water volume is within its limits. b. In accordance with the Surveillance Frequency Control Program by verifying CSP temperature when the RAB air temperature is less than 55°F or greater than 100°F. *In MODE 4, the CSP shall be OPERABLE with a minimum contained volume of at least 11 % indicated level. WATERFORD -UNIT 3 3/4 7-6 AMENDMENT NO. 137, 1QQ, 249 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT AND ANALYSIS 1. 2. Gross Activity Determination Isotopic Analysis for DOSE EQUIVALENT 1-131 Concentration WATERFORD -UNIT 3 3/4 7-8 SAMPLE AND ANALYSIS FREQUENCY In accordance with the Surveillance Frequency Control Program a) 1 per 31 days, whenever the gross activity tion indicates iodine concentrations greater than 10% of the allowable limit. b) 1 per 6 months, whenever the gross activity determination indicates iodine tions below 10% of the allowable limit. AMENDMENT NO. 249 PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES (MSIVs) LIMITING CONDITION FOR OPERATION 3.7.1.5 Two MSIVs shall be OPERABLE. APPLICABILITY: MODE 1, and MODES 2, 3, and 4, except when all MSIVs are closed and deactivated. ACTION: MODE1 With one MSIV inoperable, restore the valve to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . MODES 2, 3 and 4 With one MSIV inoperable, close the valve within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and verify the valve is closed once per 7 days. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS Note: Required to be performed for entry into MODES 1 and 2 only. 4.7.1.5 Each MSIV shall be demonstrated OPERABLE: a. By verifying full closure within 8.0 seconds when tested pursuant to the lnservice Testing Program. b. By verifying each MSIV actuates to the isolation position on an actual or simulated actuation signal in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 7-9 AMENDMENT NO. 76, 189, 190, 199, 249 PLANT SYSTEMS MAIN FEEDWATER ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.6 Each Main Feedwater Isolation Valve (MFIV) shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: Note: Separate Condition entry is allowed for each valve. With one or more MFIV inoperable, close and deactivate, or isolate the inoperable valve within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verify inoperable valve closed and deactivated or isolated once every 7 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 do not apply. SURVEILLANCE REQUIREMENTS 4. 7 .1.6 Each main feedwater isolation valve shall be demonstrated OPERABLE: a. By verifying isolation within 6.0 seconds when tested pursuant to the lnservice Testing Program. b. By verifying actuation to the isolation position on an actual or simulated actuation signal in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 7-9a AMENDMENT 189, 199, 249 3/4. 7 PLANT SYSTEMS 3/4.7.1.7 ATMOSPHERIC DUMP VALVES LIMITING CONDITION FOR OPERATION 3.7.1.7 Each Atmospheric Dump Valve (ADV) shall be OPERABLE*. APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: a. With the automatic actuation channel for one ADV inoperable, restore the inoperable ADV to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or reduce power to less than or equal to 70% RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . b. With the automatic actuation channels for both ADVs inoperable, restore one ADV to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or reduce power to less than or equal to 70% RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . c. With one ADV inoperable, for reasons other than above, restore the ADV to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 are not applicable provided one ADV is OPERABLE. SURVEILLANCE REQUIREMENTS 4.7.1.7 The ADVs shall be demonstrated OPERABLE: a. By performing a CHANNEL CHECK in accordance with the Surveillance Frequency Control Program when the automatic actuation channels are required to be OPERABLE. b. By verifying each ADV automatic actuation channel is in automatic with a setpoint of less than or equal to 1040 psia in accordance with the Surveillance Frequency Control Program when the automatic actuation channels are required to be OPERABLE. c. By verifying one complete cycle of each ADV when tested pursuant to the lnservice Testing Program. *ADV automatic actuation channels (one per ADV, in automatic with a setpoint of less than or equal to 1040 psia) are not required to be OPERABLE when less than or equal to 70% RATED THERMAL POWER for greater than 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . WATERFORD -UNIT 3 3/4 7-9b AMENDMENT NO. WS, 249 3/4.7 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued) d. By performing a CHANNEL CALIBRATION of each ADV automatic actuation channel in accordance with the Surveillance Frequency Control Program. e. By verifying actuation of each ADV to the open position on an actual or simulated automatic actuation signal in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 7-9c AMENDMENT NO 249 PLANT SYSTEMS 314.7.3 COMPONENT COOLING WATER AND AUXILIARY COMPONENT COOLING WATER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water and associated auxiliary component cooling water trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one component cooling water and associated auxiliary component cooling water train OPERABLE, restore at least two trains to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.7.3 Each component cooling water and associated auxiliary component cooling water train shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. b. In accordance with the Surveillance Frequency Control Program by verifying that each automatic valve servicing safety-related equipment actuates to its correct position on SIAS and CSAS test signals. c. In accordance with the Surveillance Frequency Control Program by verifying that each component cooling water and associated auxiliary component cooling water pump starts automatically on an SIAS test signal. WATERFORD -UNIT 3 314 7-11 AMENDMENT NO. 2-09, 249 PLANT SYSTEMS LIMITING CONDITION FOR OPERATION <Continued) ACTION: (Continued) c. With a Tornado Watch in effect, all 9 OCT fans under the missile protected portion of the OCT shall be OPERABLE. If the number of . fans OPERABLE is less than required, restore the inoperable fan(s) to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . d. With any UHS fan inoperable, determine the outside ambient temperature at least once every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and verify that the minimum fan requirements of Table 3.7-3 are satisfied (required only if the associated UHS is OPERABLE). SURVEILLANCE REQUIREMENTS 4.7.4. Each train of UHS shall be determined OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying the average water temperature and water level to be within specified limits. b. In accordance with the Surveillance Frequency Control Program, by verifying that each wet tower and dry tower fan that is not already running, starts and operates for at least 15 minutes. WATERFORD -UNIT 3 3/4 7-13 AMENDMENT NO. 95, 123, 208, 249 PLANT SYSTEMS ACTION (Continued): e. With one or more control room emergency air filtration trains inoperable due to an inoperable control room envelope boundary in MODES 5 or 6, or during load movements with or over irradiated fuel assemblies, immediately suspend load movements with or over irradiated fuel assemblies and operations involving CORE AL TERA TIONS. f. With two control room emergency air filtration trains inoperable in MODES 1, 2, 3, or 4 for reasons other than ACTION b, immediately enter LCO 3.0.3. g. With two control room emergency air filtration trains inoperable in MODES 5 and 6 or during load movements with or over irradiated fuel assemblies, immediately suspend load movements with or over irradiated fuel assemblies and operations involving CORE ALTERATIONS. SURVEILLANCE REQUIREMENTS 4.7.6.1 Each control room air filtration train (S-8) shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 1 O continuous hours with the heaters on. b. In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by: 1. Verifying that the filtration train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 4225 cfm +/- 10%. Note 1: The control room envelope (CRE) boundary may be opened intermittently under administrative control. WATERFORD -UNIT 3 3/4 7-16a AMENDMENT NO. 218, 235, 249 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 70%. 3. Verifying a system flow rate of 4225 cfm +/- 10% during train operation when tested in accordance with ANSI N510-1975. c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 70%. d. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7.8 inches water gauge while operating the train at a flow rate of 4225 cfm +/- 10%. 2. Verifying that on a safety injection actuation test signal or a high radiation test signal, the train automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks and the normal outside airflow paths isolate. 3. Verifying that heaters dissipate 10 +1.0, -1.0 kW when tested in accordance with ANSI N510-1975. 4. Verifying that on a toxic gas detection signal, the system automatically switches to the isolation mode of operation. e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 4225 cfm +/- 10%. f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 4225 cfm +/- 10%. g. Perform required control room envelope unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program. WATERFORD -UNIT 3 3/4 7-17 AMENDMENT NO. 115, 170, 194, 218, 249 PLANT SYSTEMS CONTROL ROOM AIR TEMPERATURE-OPERATING LIMITING CONDITION FOR OPERATION 3.7.6.3 Two independent control room air conditioning units shall be OPERABLE. APPLICABILITY*: MODES 1, 2, 3, and 4. ACTION: a. With one control room air conditioning unit inoperable, restore the inoperable unit to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With two control room air conditioning units inoperable, return one unit to an OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.7.6.3 Each control room air conditioning unit shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying that the operating control room air conditioning unit is maintaining average control room air temperature less than or equal to 80°F. b. At least quarterly, if not performed within the last quarter, by verifying that each control room air conditioning unit starts and operates for at least 15 minutes. *During load movements with or over irradiated fuel assemblies, TS 3.7.6.4 is also applicable. WATERFORD-UNIT 3 3/4 7-18a AMENDMENT NO. 115, 14g, 188, 218, 235, 249 PLANT SYSTEMS 3/4.7.7 CONTROLLED VENTILATION AREA SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent controlled ventilation systems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one controlled ventilation area system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.7.7 Each controlled ventilation area system shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months continuous with the heaters on. b. In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by: 1. Verifying that the controlled ventilation area system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm +/- 10%. 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 70%. 3 Verifying a system flow rate of 3000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975. WATERFORD -UNIT 3 3/4 7-19 AMENDMENT NO . .:t Q, 249 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and a relative humidity of 70%. d. In accordance with the Surveillance Frequency Control Program by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7 .8 inches \'."ater gauge while operating the system at a flow rate of 3000 cfm +/- 10%. 2. Verifying that the system starts on a Safety Injection Actuation Test Signal and achieves and maintains a negative pressure of 0.25 inch water gauge within 45 seconds. 3. Verifying that the filter cooling bypass valves can be manually cycled. 4. Verifying that the heaters dissipate 20 + 2.0, -2.0 kW when tested in accordance with ANSI N510-1975. e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 3000 cfm +/- 10%. f. After each complete or partial replacement of a charcoal absorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 3000 cfm +/- 10%. WATERFORD -UNIT 3 3/4 7-20 AMENDMENT NO. -t7-0, 194, 219, 249 Next Page is 3/4 7-27 PLANT SYSTEMS 3/4.7.12 ESSENTIAL SERVICES CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION 3. 7.12 Two independent essential services chilled water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: With only one essential services chilled water loop OPERABLE, restore two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.7.12.1 Each of the above required essential services chilled water loop shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. b. In accordance with the Surveillance Frequency Control Program by verifying that the water outlet temperature is :0 42 ° F at a flow rate of 2 500 gpm. c. Deleted d. In accordance with the Surveillance Frequency Control Program, by verifying that each essential services chilled water pump and compressor starts automatically on a safety injection actuation test signal. 4.7.12.2 The backup essential services chilled water pump and chiller shall be demonstrated OPERABLE in accordance with Specification 4.7 .12.1 whenever it is functioning as part of one of the required essential services chilled water loops. WATERFORD -UNIT 3 3/4 7-43 249 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1 E distribution system shall be: a. Determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments, indicated power availability, and b. Demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by transferring manually and automatically unit power supply from the normal circuit to the alternate circuit. 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE*: a. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the fuel level in the diesel oil feed tank, 2. Deleted, 3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the diesel oil feed tank, 4. Verifying the diesel starts**. The generator voltage and frequency shall be at least 3920 volts and 58.8 Hz in 10 seconds after the start signal. The steady state voltage and frequency shall be maintained at 4160 + 420, -240 volts and 60 +/- 1.2 Hz. The diesel generator shall be started for this test by using one of the following signals: a) Manual. b) Simulated loss-of-offsite power by itself. c) Simulated loss-of-offsite power in conjunction with an ESF actuation test signal. d) An ESF actuation test signal by itself. *All planned starts for the purpose of surveillance in this section may be preceded by a prelube period as recommended by the manufacturer. **A modified diesel generator start involving idling and gradual acceleration to synchronous speed may be used for this surveillance requirement as recommended by the manufacturer. When modified start procedures are not used, the time, speed, voltage, and frequency tolerances of this surveillance requirement must be met. WATERFORD -UNIT 3 3/4 8-3 AMENDMENT NO. 23,74,126,216, 249 ELECTRICAL POWER SYSTEM SURVEILLANCE REQUIREMENTS <Continued) 5. Verifying the generator is synchronized, loaded to an indicated 4000-4400 Kw* in accordance with the manufacturer's recommendation and operates for at least an additional 60 minutes#, and 6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses. b. In accordance with the Surveillance Frequency Control Program and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the diesel oil feed tanks. c. Deleted *This band is meant as guidance to avoid routine overloading of the engine. Loads in excess of this band for special testing under direct monitoring of the manufacturer or momentary variation due to changing bus loads shall not invalidate the test. #This surveillance requirement shall be preceded by and immediately follow without shutdown a successful performance of 4.8.1.1.2a.4 or 4.8.1.1.2d. WATERFORD -UNIT 3 3/4 8-4 AMENDMENT NO. 4,23,92,126,180,216, 249 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) d. In accordance with the Surveillance Frequency Control Program a diesel generator fast start test shall be performed in accordance with TS 4.8.1.1.2a.4. Performance of the fast start test satisfies the testing requirements specified in TS 4.8.1.1.2a.4. e. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the generator capability to reject a load of greater than or equal to 498 kW while maintaining voltage at 4160 +420, -240 volts and frequency at 60 +4.5, -1.2 Hz. 2. Verifying the generator capability to reject a load of an indicated 4000-4400 kW without tripping. The generator voltage shall not exceed 5023 volts during and following the load rejection. 3. During shutdown, simulating a loss-of-offsite power by itself, and: a) Verifying deenergization of the emergency busses and load shedding from the emergency busses. b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses and the permanently connected loads within 10 seconds after the auto-start signal, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +420, -240 volts and 60 +1.2, -0.3 Hz during this test. 4. Verifying that on an SIAS actuation test signal (without loss-of-offsite power) the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The steady-state generator voltage and frequency shall be 4160 +420, -240 volts and 60 +/- 1.2 Hz within 10 seconds after the auto-start signal; the generator voltage and frequency shall be maintained within these limits during this test. WATERFORD -UNIT 3 3/4 8-5 AMENDMENT NO. 4, 23, 74, 88, 98, 129, HW, 219, 249 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) 8. During shutdown, verifying the diesel generator's capability to: a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status. 9. During shutdown, verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizes the emergency loads with offsite power. 10. Verifying that each fuel transfer pump transfers fuel to its associated diesel oil feed tank by taking suction from the opposite train fuel oil storage tank via the installed cross connect. 11. During shutdown, verifying that the automatic load sequence timer is OPERABLE with the time of each load block within +/- 10% of the sequenced load block time. 12. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required: \ a) turning gear engaged b) emergency stop c) loss of D.C. control power d) governor fuel oil linkage tripped f. Deleted g. In accordance with the Surveillance Frequency Control Program or after any modifications which could affect diesel generator interdependence by starting the diesel generators simultaneously, during shutdown, and verifying that the diesel generators accelerate to at least 600 rpm (60 +/- 1.2 Hz) in less than or equal to 10 seconds. h. Deleted WATERFORD -UNIT 3 3/4 8-6a AMENDMENT NO. 23, 92, 126, 180, 211,216,249 ELECTRICAL POWER SYSTEMS DIESEL FUEL OIL LIMITING CONDITION FOR OPERATION 3.8.1.3 The stored diesel fuel oil shall be within limits for each required diesel generator (DG). APPLICABILITY: When associated DG is required to be OPERABLE. ACTION: (Note: Separate ACTION entry is allowed for each DG.) a. With the fuel oil storage tank volume less than 39,300 gallons and greater than 37,000 gallons, restore fuel oil storage tank volume to greater than or equal to 39,300 gallons within 5 days (provided replacement fuel oil is onsite within the first 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ). b. With one or more DGs with stored fuel oil total particulates not within limits, restore fuel oil total particulates to within limits within 7 days. c. With one or more DGs with new fuel oil properties not within limits, restore stored fuel oil properties to within limits within 30 days. d. If any of the above ACTIONS cannot be met, or if the diesel fuel oil is not within limits for reasons other than the above ACTIONS, immediately declare the associated DG(s) inoperable. SURVEILLANCE REQUIREMENTS 4.8.1.3.1 In accordance with the Surveillance Frequency Control Program verify each fuel oil storage tank volume. 4.8.1.3.2 Verify fuel oil properties of new or stored fuel oil are tested in accordance with, and maintained within the limits of, the Diesel Fuel Oil Testing Program. WATERFORD -UNIT 3 3/4 8-8a AMENDMENT NO. 249 ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum the following D.C. electrical sources shall be OPERABLE: a. 125-volt Battery Bank No. 3A-S and one associated full capacity charger (3A 1-S or 3A2-S). b. 125-volt Battery Bank No. 3B-S and one associated full capacity charger (3B1-S or 3B2-S). c. 125-volt battery Bank No. 3AB-S and one associated full capacity charger (3AB1-S or 3AB2-S). APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. With one of the required battery banks inoperable, restore the inoperable battery bank to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With one of the required full capacity chargers inoperable, demonstrate the OPERABILITY of its associated battery bank by performing Surveillance Requirement 4.8.2.1 a.1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If any Category A limit in Table 4.8-2 is not met, declare the battery inoperable. SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and at least one associated charger shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program by verifying that: 1. The parameters in Table 4.8-2 meet the Category A limits, and 2. The total battery terminal voltage is greater than or equal to 125 volts on float charge. WATERFORD -UNIT 3 314 8-9 AMENDMENT NO. 249 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) b. In accordance with the Surveillance Frequency Control Program and within 7 days after a battery discharge with battery terminal voltage below 110-volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that: 1. The parameters in Table 4.8-2 meet the Category B limits, 2. There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 1 o-6 ohms, and 3. The average electrolyte temperature of a random sample of at least ten of the connected cells is above 70°F. c. In accordance with the Surveillance Frequency Control Program by verifying that: 1. The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, 2. The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material, 3. The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-5 ohms, and 4. The battery charger will supply at least 150 amperes for 3A 1-S, 3A2-S, 381-S and 382-S and 200 amperes for 3AB1-S and 3AB2-S at greater than or equal to 132 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . d. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test. e. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This performance discharge test may be performed in lieu of the battery service test required by Surveillance Requirement 4.8.2.1d. f. Annual performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating. WATERFORD -UNIT 3 3/4 8-10 AMENDMENT N0.,...+.7, 249 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION <Continued) ACTION: a. With one of the required divisions of A.C. ESE busses not fully energized, reenergize the division within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. With one A.C. SUPS bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C. bus: (1) reenergize the A.C. SUPS bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months and (2) reenergize the A.C. SUPS bus from its associated inverter connected to its associated D.C. bus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . c. With one D.C. bus not connected to its associated battery bank, reconnect the D.C. bus from its associated OPERABLE battery bank within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined energized in the required manner in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses. WATERFORD -UNIT 3 3/4 8-14 AMENDMENT NO. 249 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner: a. One division of A.C. ESE busses consisting of one 4160 volt and one 480-volt A.C. ESE bus (3A3-S and 3A31-S or 3B3-S and 3B31-S). b. Two 120-volt A.C. SUPS busses energized from their associated inverters connected to their respective D.C. busses (3MA-S, 3MB-S, 3MC-S, or 3MD-S). c. One 120-volt A.C. SUPS Bus (3A-S or 3B-S) energized from its associated inverter connected to its respective D. C. bus. d. One 125-volt D.C. bus (3A-DC-S or 3B-DC-S) connected to its associated battery bank. APPLICABILITY: MODES 5 and 6. ACTION: With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or boron concentration, or load movements with or over irradiated fuel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible. SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses. WATERFORD -UNIT 3 3/4 8-15 AMENDMENT NO. 185, 235, 249 ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 Primary and backup containment penetration conductor overcurrent protective devices associated with each containment electrical penetration circuit shall be OPERABLE. The scope of these protective devices excludes those circuits for which credible fault currents would not exceed the trical penetration design rating. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. With one or more of the above required containment penetration conductor overcurrent devices inoperable: 1. Restore the protective device(s) to OPERABLE status or deenergize the circuit(s) by tripping, racking out, or removing the alternate device or racking out or removing the inoperable device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and 2. Declare the affected system or component inoperable, and 3. Verify at least once per 7 days thereafter the alternate device is tripped, racked out, or removed, or the device is racked out or removed. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . b. The provisions of Specification 3.0.4 are not applicable to overcurrent devices which have the inoperable device racked out or removed or, which have the alternate device tripped, racked out, or removed. SURVEILLANCE REQUIREMENTS 4.8.4.1 The above noted primary and backup containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE: a. In accordance with the Surveillance Frequency Control Program: 1. By verifying that the medium voltage (4-15 kV) circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers of each voltage level, and performing the following: (a) A CHANNEL CALIBRATION of the associated protective relays, and (b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed. WATERFORD -UNIT 3 3/4 8-16 AMENDMENT NO. +a, 249 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) (c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. 2. By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers, except for those breakers with external trip devices,* shall consist of injecting a current in excess of the breakers' nominal setpoint and measuring the response time. The measured response time will be compared to the manufacturer's data to ensure that it is less than or equal to a value specified by the manufacturer. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. b. In accordance with the Surveillance Frequency Control Program by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations. *Testing of these circuit breakers (i.e., the 480 volts power from low voltage switchgear) shall be performed in accordance with the manufacturer's recommendations. WATERFORD -UNIT 3 3/4 8-17 AMENDMENT NO. 249 ELECTRICAL POWER SYSTEMS MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.2 The thermal overload protection and bypass devices, integral with the motor starter, of each valve used in safety systems shall be OPERABLE. APPLICABILITY: Whenever the motor operated valve is required to be OPERABLE. ACTION: With one or more of the thermal overload protection and/or bypass devices inoperable, declare the affected valve(s) inoperable and apply the appropriate ACTION Statement(s) for the affected valve(s). SURVEILLANCE REQUIREMENTS 4.8.4.2 The above required thermal overload protection and bypass devices shall be demonstrated OPERABLE. a. In accordance with the Surveillance Frequency Control Program, by the performance of a CHANNEL FUNCTIONAL TEST of the bypass circuitry for those thermal overload devices which are either: 1. Continuously bypassed and temporarily placed in force only when the valve motors are undergoing periodic or maintenance testing, or 2. Normally in force during plant operation and bypassed under accident conditions. b. In accordance with the Surveillance Frequency Control Program by the performance of a CHANNEL CALIBRATION of a representative sample of at least 25% of: 1. All thermal overload devices which are not bypassed, such that each nonbypassed device is calibrated in accordance with the Surveillance Frequency Control Program. 2. All thermal overload devices which are continuously bypassed and temporarily placed in force only when the valve motors are undergoing periodic or maintenance testing, and thermal overload devices normally in force and bypassed under accident conditions such that each thermal overload is calibrated and each valve is cycled through at least one complete cycle of full travel with the motor-operator when the thermal overload is OPERABLE and not bypassed, in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 8-52 AMENDMENT NO. 249 Corros#oR Jotter of 10 5 92 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the reactivity conditions specified in the COLR is met. APPLICABILITY: MODE 6*. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE AL TERA TIONS or positive reactivity changes and initiate action to restore boron concentration to within COLR limits. SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: a. Removing or unbolting the reactor vessel head, and b. Withdrawal of any CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel. 4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis in accordance with the Surveillance Frequency Control Program. *The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed. WATERFORD -UNIT 3 3/4 9-1 AMENDMENT NO. 102,129, 182, 249 REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE and operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room. APPLICABILITY: MODE 6. ACTION: a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE AL TERA TIONS or operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of Technical Specification 3.9.1. b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of: a. A CHANNEL CHECK in accordance with the Surveillance Frequency Control Program, b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS, and c. A CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 3/4 9-2 AMENDMENT NO. 249 REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status: a. The equipment door is closed, b. A minimum of one door in each airlock is capable of being closed, and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either: 1. Closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. Capable of being closed by an OPERABLE containment purge and exhaust isolation system. Note: Penetration flow path(s) described in a, b, and c above, that provides direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls. APPLICABILITY: During CORE AL TERA TIONS or load movements with or over irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or load movements with or over irradiated fuel in the containment building. SURVEILLANCE REQUIREMENTS 4.9.4.1 Verify each required containment penetration is in the required status prior to the start of and in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS or load movements with or over irradiated fuel within containment. 4.9.4.2 Verify each required containment purge and exhaust valve actuates to the isolation position on an actual or simulated actuation signal in accordance with the Surveillance Frequency Control Program or load movements with or over irradiated fuel within containment. NOTE -SR 4.9.4.2 is not required to be met for containment purge and exhaust valve(s) in penetrations closed to comply with LCO 3.9.4.c.1. WATERFORD -UNIT 3 3/4 9-4 AMENDMENT NO. 169, 2a1, 2a5, 249 REFUELING OPERATIONS 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one shutdown cooling train shall be OPERABLE and in operation.* APPLICABILITY: MODE 6 when the water level above the top of the fuel seated in the reactor pressure vessel is greater than or equal to 23 feet. ACTION: With no shutdown cooling train OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of Technical Specification 3.9.1 and immediately initiate corrective action to return the required shutdown cooling train to OPERABLE and operating status. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm** in accordance with the Surveillance Frequency Control Program. *The shutdown cooling loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8-hour period during the performance of CORE AL TERA TIONS in the vicinity of the reactor pressure vessel hot legs, provided no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the minimum required boron concentration of Technical Specification 3.9.1. **The minimum flow may be reduced to 3000 gpm after the reactor has been shut down for greater than or equal to 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months or by verifying at least once per hour that the RCS temperature is less then 135°F. The minimum flow may be reduced to 2000 gpm after the reactor has been shut down for greater than or equal to 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months . WATERFORD -UNIT 3 3/4 9-8 AMENDMENT NO. 35, 148, 185, 249 REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION # 3.9.8.2 Two independent shutdown cooling trains shall be OPERABLE and at least one shutdown cooling train shall be in operation.* APPLICABILITY: MODE 6 when the water level above the top of the fuel seated in the reactor pressure vessel is less than 23 feet. ACTION: a. With one of the required shutdown cooling trains inoperable, immediately initiate corrective action to return the required train to OPERABLE status, or to establish greater than or equal to 23 feet of water above the top of the fuel seated in the reactor pressure vessel. b. With no shutdown cooling train OPERABLE and in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of Technical Specification 3.9.1 and immediately initiate corrective action to return the required shutdown cooling train to OPERABLE and operating status. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm** in accordance with the Surveillance Frequency Control Program. #Only one shutdown cooling train is required to be OPERABLE and in operation provided there are no irradiated fuel assemblies seated within the reactor pressure vessel. *The shutdown cooling loop may be removed from operations for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs, provided no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the minimum required boron concentration of Technical Specification 3.9.1. **The minimum flow may be reduced to 3000 gpm after the reactor has been shut down for greater than or equal to 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months or by verifying at least once per hour that the RCS temperature is less than 135°F. The minimum flow may be reduced to 2000 gpm after the reactor has been shut down for greater than or equal to 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months . WATERFORD -UNIT 3 3/4 9-9 AMENDMENT NO. 35, 148, 185, 249 REFUELING OPERATIONS 3/4.9.10 WATER LEVEL -REACTOR VESSEL FUEL ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.9.10.1 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange. APPLICABILITY: During movement of fuel assemblies within the reactor pressure vessel when either the fuel assemblies being moved or the fuel assemblies seated within the reactor pressure vessel are irradiated. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies within the pressure vessel. SURVEILLANCE REQUIREMENTS 4.9.10.1 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program thereafter during movement of fuel assemblies. WATERFORD -UNIT 3 314 9-11 AMENDMENT NO. 249 REFUELING OPERATIONS LIMITING CONDITION FOR OPERATION 3.9.10.2 At least 23 feet of water shall be maintained over the top of the fuel seated in the reactor pressure vessel. APPLICABILITY: During movement of CEAs within the reactor pressure vessel, when the fuel assemblies seated within the reactor pressure vessel are irradiated. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of CEAs within the pressure vessel. SURVEILLANCE REQUIREMENTS 4.9.10.2 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program thereafter during movement of CEAs. WATERFORD -UNIT 3 3/4 9-12 AMENDMENT NO. 249 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL -SPENT FUEL POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool. ACTION: With the requirement of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the spent fuel pool shall be determined to be at least its minimum required depth in accordance with the Surveillance Frequency Control Program when irradiated fuel assemblies are in the spent fuel pool. WATERFORD -UNIT 3 3/4 9-13 AMENDMENT NO. 249 REFUELING OPERATIONS 3/4.9.12 SPENT FUEL POOL (SFP) BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.12 The spent fuel pool boron concentration shall 1900 ppm. APPLICABILITY: When fuel assemblies are stored in the SFP. ACTION: a. With the spent fuel pool boron concentration not within limits immediately suspend movement of fuel in the SFP and immediately initiate actions to restore boron concentration to within limits. b. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.12 Verify the spent fuel pool concentration is within limits in accordance with the Surveillance Frequency Control Program. WATERFORD -UNIT 3 314 9-13a AMENDMENT NO. 249 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 or 3.1.1.2 may be suspended for measurement of CEA worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s). APPLICABILITY: MODES 2 AND 3*. ACTION: a. With any CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate boration to restore the SHUTDOWN MARGIN required by Specification 3.1.1.1. b. With all CEAs fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate boration to restore the SHUTDOWN MARGIN required by Specification 3.1.1.2. SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each CEA required either partially or fully withdrawn shall be determined in accordance with the Surveillance Frequency Control Program. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. *Operation in MODE 3 shall be limited to 6 consecutive hours. WATERFORD -UNIT 3 3/4 10-1 AMENDMENT NO. 11,141, 182, 249 SPECIAL TEST EXCEPTIONS 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT. INSERTION. AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided: a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below. APPLICABILITY: MODES 1 and 2. ACTION: With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended, either: a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended and shall be verified to be within the test power plateau. 4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the lncore Detection Monitoring System pursuant to the requirements of Specifications 4.2.1.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended. WATERFORD -UNIT 3 3/4 10-2 Amendment No. 13 136, 182, 249 SPECIAL TEST EXCEPTIONS 3/4.10.3 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.3 The noted requirements of Tables 2.2-1 and 3.3-1 may be suspended during the performance of startup and PHYSICS TESTS, provided: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and either b. The reactor trip setpoints of the OPERABLE power level channels are set at less than or equal to 20% of RA TED THERMAL POWER, or c. The core protection calculator operating bypass permissive setpoints are increased to greater than the logarithmic power-hi trip setpoint specified in Table 2.2-1 and less than 5% RATED THERMAL POWER. APPLICABILITY: During startup and PHYSICS TESTS. ACTION: With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately trip the reactor. SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during startup and PHYSICS TESTS. 4.10.3.2 Each wide range logarithmic and power level neutron flux monitoring channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating startup and PHYSICS TESTS. WATERFORD -UNIT 3 3/4 10-3 AMENDMENT NO . .+4-, 249 SPECIAL TEST EXCEPTIONS 3/4.10.5 NATURAL CIRCULATION TESTING LIMITING CONDITION FOR OPERATION 3.10.5 The limitation of Specification 3.4.1.2 may be suspended during the performance of natural circulation testing, provided the Reactor Coolant System saturation margin is maintained greater than or equal to 20°F. APPLICABILITY: MODE 3 during natural circulation testing. ACTION: With the Reactor Coolant System saturation margin less than 20°F, immediately place at least one reactor coolant loop in operation, with at least one reactor coolant pump. SURVEILLANCE REQUIREMENTS 4.10.5.1 The saturation margin shall be determined to be within the above limits by continuous monitoring with the saturation margin monitors required by Table 3.3-10 or, by calculating the saturation margin in accordance with the Surveillance Frequency Control Program. 4.10.5.2 The saturation margin monitor shall be demonstrated OPERABLE by performance of a CHANNEL CHECK within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to initiating natural circulation testing. WATERFORD -UNIT 3 3/4 10-5 AMENDMENT NO. 249 RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION OR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 8.5 x 104 curies noble gases (considered as Xe-133 equivalent). APPLICABILITY: At all times. ACTION: a. With the quantity of radioactive material in any gas storage tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limits and describe the events leading to this condition in the next Radioactive Effluent Release Report, pursuant to Specification 6.9.1.8. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank on-service shall be determined to be within the above limit in accordance with the Surveillance Frequency Control Program until the quantity exceeds 4.25 x 104 curies noble gases (50% of allowed limit) and then at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank. Tanks isolated for decay will be sampled to verify above limit is met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following removal from service. WATERFORD-UNIT 3 3/411-17 AMENDMENT NO 249 CorreetioR .ieuer of 5 16 90 ADMINISTRATIVE CONTROLS 6.5.17 6.5.18 Control Room Envelope Habitability Program (Continued) a. The definition of the CRE and the CRE boundary. b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance. c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0. d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the control room emergency air filtration, operating at the flow rate required by SR 4.7.6.1.b, in accordance with the Surveillance Frequency Control Program. The results shall be trended and used as part of the assessment of the CRE boundary. e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air in leakage measured by the testing described in paragraph c. The unfiltered air in leakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air in leakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis. f. The provisions of SR 4.0.2 are applicable to the FREQUENCIES for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively. Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program. b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-informed Method for Control of Surveillance Frequencies, "Revision 1. c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. WATERFORD -UNIT 3 6-9 AMENDMENT NO. 249 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 249 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS. INC. WATERFORD STEAM ELECTRIC STATION. UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By letter dated June 17, 2015 (Reference 1), as supplemented by letters dated March 3, 2016 (Reference 2), April 28, 2016 (Reference 3), and July 12, 2016 (Reference 21 ), Entergy Operations, Inc. (the licensee), submitted a license amendment request (LAR) which proposed changes to the Technical Specifications (TSs) for Waterford Steam Electric Station, Unit 3 (Waterford 3), which are contained in Appendix A of Renewed Facility Operating License No. NPF-38. The licensee requested to revise the Waterford 3 TSs by relocating specific surveillance requirement (SR) frequencies to a licensee-controlled program. The licensee requested to revise the TSs to require that changes to such surveillance frequencies will be made in accordance with Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative Sb, Risk-Informed Method for Control of Surveillance Frequencies," Revision 1, dated April 2007 (Reference 4). The requested change is the adoption of U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Standard Technical Specifications (STSs) Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed TSTF] Initiative 5b" (Reference 5). The Federal Register (FR) notice published on July 6, 2009 (74 FR 31996), announced the availability of TSTF-425, Revision 3. By letter dated January 22, 2016 (Reference 6), the NRC sent a request for additional information (RAI) to the licensee. By letter dated March 3, 2016, the licensee responded to this request. As a consequence of this response, the NRC staff sent another RAI by letter dated April, 12, 2016 (Reference 7). By letter dated April 28, 2016, the licensee responded to this second request. By letter dated July 12, 2016, the licensee provided a supplement proposing minor edits to the Technical Specifications. The supplements from the licensee dated March 3, 2016, April 28, 2016, and July 12, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change Enclosure 2 the NRC staff's original proposed no significant hazards consideration (NSHC) determination as published in the Federal Register on September 1, 2015 (80 FR 52805).

2.0 REGULATORY EVALUATION

2.1 Description of the Proposed Changes The licensee proposed to modify the Waterford 3 TSs by relocating specific surveillance frequencies to a licensee-controlled program (i.e., the Surveillance Frequency Control Program (SFCP)) in accordance with NEI 04-10, Revision 1. The licensee .stated that the proposed change is consistent with the adoption of NRG-approved TSTF-425, Revision 3. When implemented, TSTF-425, Revision 3, relocates most periodic frequencies of TS surveillances to the SFCP, and provides requirements for the new program in the Administrative Controls sections of the TSs. All surveillance frequencies can be relocated except the following:

Frequencies that reference other approved programs for the specific interval, such as the lnservice Testing Program or the Primary Containment Leakage Rate Testing Program;

Frequencies that are purely event-driven (e.g., "each time the control rod is withdrawn to the 'full out' position");

Frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching 2:: [greater than or equal to] 95% RTP [rated thermal power]"); and

Frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of an SR (e.g., "drywell to suppression chamber differential pressure decrease"). The SFCP describes the requirements for the program to control changes to the relocated surveillance frequencies. The licensee proposed to add the SFCP to Section 6.0, "Administrative Controls," of the Waterford 3 TSs. The TS Bases for each affected surveillance would be revised to state that the frequency is controlled under the SFCP. The current information in the TS Bases will be relocated to the licensee-controlled SFCP. The proposed changes to the Administrative Controls sections of the TSs to incorporate the SFCP include a specific reference to NEI 04-10, Revision 1, as the basis for making any changes to the surveillance frequencies once they are relocated out of the TSs. By letter dated September 19, 2007 (Reference 8), the NRC staff approved Topical Report NEI 04-10, Revision 1, as acceptable for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, Revision 1, and the safety evaluation (SE) providing the basis for NRC acceptance of NEI 04-10, Revision 1. For Waterford 3, the licensee proposed other changes and deviations from TSTF-425, which are discussed in Section 3.3 of this SE.

2.2 Applicable Commission Policy Statements In the "Final Policy Statement: Technical Specifications for Nuclear Power Plants," dated July 22, 1993 (58 FR 39132), the NRC addressed the use of probabilistic safety analysis (PSA, currently referred to as probabilistic risk assessment or PRA) in STSs. In this 1993 publication, the NRC states, in part: The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria [of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36) to be deleted from Technical Specifications based solely on PSA (Criterion 4). However, if the results of PSA indicate that Technical Specifications can be relaxed or removed, a deterministic review will be performed. The Commission Policy in this regard is consistent with its Policy Statement on "Safety Goals for the Operation of Nuclear Power Plants," 51 FR 30028, published on August 21, 1986. The Policy Statement on Safety Goals states in part, "* *

probabilistic results should also be reasonably balanced and supported through use of deterministic arguments. In this way, judgments can be made * *

about the degree of confidence to be given these [probabilistic] estimates and assumptions. This is a key part of the process for determining the degree of regulatory conservatism that may be warranted for particular decisions. This defense-in-depth approach is expected to continue to ensure the protection of public health and safety." The Commission will continue to use PSA, consistent with its policy on Safety Goals, as a tool in evaluating specific line-item improvements to Technical Specifications, new requirements, and industry proposals for risk-based Technical Specification changes. Approximately 2 years later, the NRC provided additional detail concerning the use of PRA in the "Final Policy Statement: Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities," dated August 16, 1995 (60 FR 42622). In this publication, the NRC states, in part: The Commission believes that an overall policy on the use of PRA methods in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that would promote regulatory stability and efficiency. In addition, the Commission believes that the use of PRA technology in NRC regulatory activities should be increased to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach .... PRA addresses a broad spectrum of initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for multiple and common cause failures. The treatment therefore goes beyond the single failure requirements in the deterministic approach. The probabilistic approach to regulation is, therefore, considered an extension and enhancement of traditional regulation by considering risk in a more coherent and complete manner. Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency. This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the scope of PRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data. Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA: ( 1) The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy. (2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised. (3) PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review. (4) The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees. 2.3 Applicable Regulations In 10 CFR 50.36, 'Technical specifications," the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; and (5) administrative controls. Items in these categories will remain in the Waterford 3 TSs. Section 50.36(c)(3) of 10 CFR states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The FR notice published on July 6, 2009 (74 FR 31996), which announced the availability of TSTF-425, Revision 3, states that the addition of the SFCP to the TSs provides the necessary administrative controls to require that surveillance frequencies relocated to the SFCP are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The FR notice also states that changes to surveillance frequencies in the SFCP are made using the methodology contained in NEI 04-10, Revision 1, as approved by NRC staff, including qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, and recommended monitoring of structures, systems, and components (SSCs), and are required to be documented. Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants" (i.e., the Maintenance Rule), and 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action, require licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. Such failures can result in the licensee increasing the frequency of a surveillance test. In addition, by specifying that the TSs require that changes to the frequencies listed in the SFCP be made in accordance with NEI 04-10, Revision 1, the licensee will be required to monitor the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs. 2.4 Applicable NRC Regulatory Guides and Review Plans Regulatory Guide (RG) 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2 (Reference 9), describes an acceptable risk-informed approach for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This regulatory guide also provides risk acceptance guidelines for evaluating the results of such evaluations. RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 1 (Reference 10), describes an acceptable risk-informed approach specifically for assessing proposed TS changes. RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 11 ), describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results such that the PRA can be used in regulatory decisionmaking for LWRs. General guidance for evaluating the technical basis for proposed risk-informed changes is provided in NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance" (Reference 12). Guidance on evaluating PRA technical adequacy is provided in SRP, Chapter 19, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load," Revision 3 (Reference 13). More specific guidance related to risk-informed TS changes is provided in SRP, Chapter 16, Section 16.1, "Risk-Informed Decisionmaking: Technical Specifications," Revision 1 (Reference 14), which includes changes to surveillance test intervals (STls) (i.e., surveillance frequencies) as part of risk-informed decisionmaking. Section 19.2 of the SRP references the same criteria as RG 1.177, Revision 1, and RG 1.17 4, Revision 2, and states that a risk-informed application should be evaluated to ensure that the proposed changes meet the following key principles:

The proposed change meets the current regulations, unless it explicitly relates to a requested exemption or rule change.

The proposed change is consistent with the defense-in-depth philosophy.

The proposed change maintains sufficient safety margins.

When proposed changes result in an increase in core damage frequency (CDF) or risk, the increase(s) should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

The impact of the proposed change should be monitored using performance measurement strategies.

3.0 TECHNICAL EVALUATION

The licensee's adoption of TSTF-425, Revision 3, provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the Administrative Controls sections of TSs. The changes to the Administrative Controls section of the TSs will also require the application of NEI 04-10, Revision 1, as approved by the NRC, for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes described in TSTF-425, Revision 3, included documentation regarding the PRA technical adequacy consistent with RG 1.200, Revision 2 (Reference 11 ). NEI 04-10, Revision 1, states, in part, that PRA methods are used with plant performance data and other considerations to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is consistent with guidance provided in RG 1.17 4, Revision 2 (Reference 9), and RG 1.177, Revision 1 (Reference 10), in support of changes to STls. 3.1 Review Methodology RG 1.177, Revision 1, identifies five key safety principles required for risk-informed changes to TSs. Each of these principles is addressed by NEI 04-10, Revision 1.

3.1.1 The Proposed Change Meets Current Regulations Unless it is Explicitly Related to a Requested Exemption or Rule Change Section 50.36(c)(3) of 10 CFR requires that TSs include surveillances, which are "requirements relating to test, calibration, or inspection to assure that necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The licensee is required by its TSs to perform surveillance tests, calibration, or inspection on specific safety-related equipment (e.g., reactivity control, power distribution, electrical, and instrumentation) to verify system operability. Surveillance frequencies are based primarily upon deterministic methods, such as engineering judgment, operating experience, and manufacturer's recommendations. The licensee's use of approved methodologies identified in NEI 04-10, Revision 1, provides a way to establish informed surveillance frequencies that complements the deterministic approach and supports the NRC's traditional defense-in-depth philosophy. The SRs themselves are remaining in the TSs, as required by 10 CFR 50.36(c)(3). This change is analogous with other NRG-approved TS changes in which the SRs are retained in TSs, but the related surveillance frequencies are relocated to licensee-controlled documents, such as surveillances performed in accordance with the lnservice Testing Program and the Primary Containment Leakage Rate Testing Program. Thus, this proposed change complies with 10 CFR 50.36(c)(3) by retaining the requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The regulatory requirements in 10 CFR 50.65 and 10 CFR Part 50, Appendix B, and the monitoring required by NEI 04-10, Revision 1, ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions taken. The licensee's SFCP ensures that SRs specified in the TSs are performed at intervals sufficient to assure the above regulatory requirements are met. In light of the above, the NRC staff concludes that the proposed change meets the first key safety principle of RG 1.177, Revision 1 (Reference 10), by complying with current regulations. 3. 1.2 The Proposed Change Is Consistent With the Defense-in-Depth Philosophy The defense-in-depth philosophy (i.e., the second key safety principle of RG 1.177, Revision 1 ), is maintained if:

A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers). (Because the scope of the proposed methodology is limited to revision of surveillance frequencies, the redundancy, independence, and diversity of plant systems are not impacted.)

Defenses against potential common-cause failures (CCFs) are preserved, and the potential for the introduction of new CCF mechanisms is assessed.

Independence of barriers is not degraded.

Defenses against human errors are preserved.

The intent of the General Design Criteria in 10 CFR Part 50, Appendix A, is maintained. The changes to the Administrative Controls section of the TSs will require the application of NEI 04-10, Revision 1, for any changes to surveillance frequencies within the SFCP. NEI 04-10, Revision 1, uses both the CDF and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. The guidance of RG 1.174, Revision 2 (Reference 9), and RG 1.177, Revision 1(Reference10), for changes to CDF and LERF, is achieved by evaluation using a comprehensive risk analysis, which assesses the impact of proposed changes including contributions from human errors and CCFs. Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of CCFs. The NRC staff concludes that both the quantitative risk analysis and the qualitative considerations assure that a reasonable balance of defense-in-depth is maintained to ensure protection of public health and safety, satisfying the second key safety principle of RG 1.177, Revision 1. 3.1.3 The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that will be conducted by the licensee under the SFCP when frequencies are revised will assess the impact of the proposed frequency change to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the proposed surveillance test frequency change is not in conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis; or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist. The design, operation, testing methods, and acceptance criteria for SSCs specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plants' licensing bases, including the Updated Final Safety Analysis Report and TS Bases, because these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. On this basis, the NRC staff concludes that safety margins are maintained by the proposed methodology, and the third key safety principle of RG 1.177, Revision 1, is satisfied.

3.1.4 When Proposed Changes Result in an Increase in CDF or Risk, the Increases Should Be Small and Consistent with the Intent of the Commission's Safety Goal Policy Statement RG 1.177, Revision 1, provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies which requires identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency, and performance of sensitivity and uncertainty evaluations. The changes to the Administrative Controls section of the TSs will require application of NEI 04-10, Revision 1, in the SFCP. NEI 04-10, Revision 1, satisfies the intent of RG 1.177, Revision 1, guidance for evaluation of the change in risk, and for assuring that such changes are small by providing the technical methodology to support risk-informed TSs for control of surveillance frequencies. 3.1.4.1 Quality of the PRA The quality of the licensee's PRA must be commensurate with the safety significance of the proposed TS change and the role the PRA plays in justifying the change. That is, the greater the change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must be employed when ensuring the quality of the PRA. RG 1.200 (Reference 11) provides regulatory guidance for assessing the technical adequacy of a PRA. The current revision (i.e., Revision 2) of this RG endorses, with clarifications and qualifications, the use of (1) American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Standard, RA-Sa-2009, "Addenda to ASME RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (i.e., the PRA Standard) (Reference 15), (2) NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision 1 (Reference 16), and (3) NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 2 (Reference 17). The licensee has performed an assessment of the PRA models used to support the SFCP using the guidance of RG 1.200, Revision 2 (with the exception of the shutdown PRA model for which an NRG-endorsed PRA standard does not exist), to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability Category (CC) II of the NRG-endorsed PRA standard is the target capability level for supporting requirements for the internal events PRA for this application. Any identified deficiencies to those requirements are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including the use of sensitivity studies where appropriate, in accordance with NEI 04-10, Revision 1. A full-scope peer review by the Westinghouse Owners Group of the Waterford 3 internal events PRA was performed in August 2009 using RG 1.200, Revision 1 and PRA standard ASME/ANS RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B to ASME RA-S-2002" (Reference 18). The licensee performed a self-assessment of the internal events PRA in 2013 using RG 1.200, Revision 2 and ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1 /Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated Februrary 2, 2009. The LAR Attachment 2, Table 3-1, "Status of Identified Gaps to Capability Category II of the ASME PRA Standard" (Reference 1), identifies PRA

-10 -standard Supporting Requirements determined by the self-assessment not to meet CC II and provides an assessment of the impact of these findings (referred to in the LAR as "gaps") for this application. The LAR explains that the impact of each "gap" will be reviewed as part of STI change evaluations. Though not provided in this LAR, the facts and observations (F&Os) or findings from the 2009 full-scope internal events PRA peer review were provided in the Waterford 3 National Fire Protection Association (NFPA) 805 transition report LAR dated November 17, 2011 (Reference 19), along with dispositions for that application. The Waterford 3 NFPA 805 LAR also provided findings from the November 2010 full-scope peer review and 2012 and 2013 followup focused-scoped peer reviews, along with their dispositions. The NRC staff reviewed the self-assessment findings for the internal events PRA, the 2009 full-scope peer review findings on the internal events PRA, and the 2010, 2012, and 2013 full-scope and focused-scope peer review findings for the fire PRA as provided in the Waterford 3 NFPA 805 LAR. The NRC staff assessed these "gaps" and findings to identify whether any incompleteness could impact this application and to ensure any deficiencies in meeting CC II can be addressed for the SFCP per the NEI 04-10, Revision 1 methodology. The NRC staff noted several self-assessment findings identified in LAR Attachment 2, Table 3-1 (Reference 1 ), associated with the internal flooding PRA that are dispositioned as having little or no impact on STI evaluations without explanation of this determination. The NRC staff requested further information in RAl-1, submitted by letter dated January 22, 2016 (Reference 6), about "gaps" that appeared to represent incompleteness in the internal flooding PRA modelling. In RAl-1.A (Reference 6), associated with Gap #1, the NRC staff noted that even though pipe break flooding sources were conservatively treated, exclusion of other flooding sources could result in underestimation of internal flooding risk and impact STI evaluations. The NRC staff, therefore, requested justification of excluding flooding sources. In response to RAl-1.A by letter dated March 3, 2016 (Reference 2), the licensee explained that "non-pipe break" flooding sources were considered in the internal flooding PRA (IFPRA) and cited examples of flooding caused by failures of valves, tanks, heat exchangers, pumps, seals, gaskets, and fitting. The licensee, however, also explained that human-induced flooding events were screened from the analysis because in such cases there would be a heightened awareness that would facilitate immediate recovery of the event. The licensee concluded that "there is no underestimation of flood frequencies." The NRC staff notes that PRA standard Supporting Requirements IFSO-A4 and IFS0-82 appear to be met because the criteria for screening the excluded flooding sources are documented. The NRC staff also notes that the internal flooding CDF and LERF reported in the licensee's response to NFPA 805 RAI dated May 14, 2015 (Reference 20), is low (i.e., 2.5E-06 and 8.6E-08, respectively). The NRC staff observes that a small increase in internal CDF and LERF values resulting from inclusion of excluded flooding sources would be well below the RG 1.17 4 limit of 1 E-06/year CDF and 1 E-07 /year LERF allowed for individual STI changes allowed by the SFCP (i.e., Step 12-A2 of NEI 04-10, Revision 1 ). Based on these observations, the NRC staff concludes that this modeling exclusion is acceptable because it has a minor effect on STI evaluations. In RAl-1.B (Reference 6), associated with Gap #2, NRC staff noted that there are IFPRA assumptions for which engineering calculations were unavailable, and therefore, requested that the licensee justify that the unsupported assumptions have no impact on the internal flooding risk that impacts STI evaluations. In its response to RAl-1.B (Reference 2), the licensee

-11 -explained that the cited engineering calculations to support the IFPRA pipe break scenarios are now available as part of the PRA documentation. The NRC staff concludes that this issue is resolved because the missing engineering calculations to support the cited assumptions are now part of the IFPRA documentation. In RAl-1.C (Reference 6), associated with Gaps #3 and #6, NRC staff noted that flooding events in the Fire Pump House and the Condensate Polisher Building were excluded from the IFPRA modeling and requested further justification given that these exclusions could contribute to underestimation of internal flooding risk that impact STI evaluations. In response to RAl-1.C (Reference 2), the licensee explained that any flooding in the Fire Pump House would not damage risk significant equipment or cause a "malfunction" that would lead the plant to trip. For the Condensate Polisher Building, the licensee explained that the worst-case flooding scenario is a plant trip with loss of main feedwater and loss of the 480 volt building switchgear. The licensee explained that the conditional core damage probability of this scenario is 4.68E-05, and therefore has a "negligible impact [on] this evaluation" when combined with the flooding frequency. The NRC staff notes again that the internal flooding CDF and LERF are low, as reported in the licensee's response to the NFPA 805 RAI (Reference 20), dated May 14, 2015, and that a small increase in internal CDF and LERF values due to inclusion of excluded flooding sources would be below the RG 1.17 4 limit of 1 E-06/year CDF and 1 E-07 /year LERF allowed for individual STI changes by the SFCP program. The NRC staff concludes that these modeling exclusions are acceptable because they would have minor impact on STI evaluations. In RAl-1.D (Reference 6), associated with Gap #4, NRC staff noted that, per the gap assessment finding, the licensee inappropriately applied a reduction factor to convert rupture flow rates to spray flow rates in the IFPRA, and requested that the licensee justify this treatment. In response to RAl-1. D (Reference 2), the licensee explained that the use of the reduction factor implied in the description in the licensee's analysis was not applied to determine a spray flow rate; rather, industry guidance was used to define spray flow rates. The NRC staff concludes that this issue is resolved because an inappropriate application of a reduction factor to convert rupture flow rates to spray flow rates was not used. In RAl-1.E (Reference 6), associated with Gap #8, NRC staff noted that numerical uncertainties were not propagated as part of IFPRA quantification, and therefore the state of knowledge correlation (SOKC) between failure rates was not considered, which might contribute to underestimation of the flooding risk that impacts STI calculations. The NRC staff, therefore, requested justification for not incorporating SOKC. In its response to RAl-1. E (Reference 2), the licensee explained that the increase in internal flooding CDF and LERF due to SOKC would be "minor" and would have a "negligible impact on STI evaluations." The NRC staff acknowledges that incorporation of SOKC would have only a small impact on the mean IFPRA CDF and LERF (e.g., might cause a fractional increase of 10 percent) and notes that in STI calculations, the SOKC (with the exception of the failure rate associated with unavailable equipment) would be the same in the baseline and variant cases. In addition, the NRC staff notes that the internal flooding CDF and LERF are low, as reported in the licensee's response to the NFPA 805 RAI (Reference 20), and that a small increase in internal CDF and LERF values that might result from consideration of SOKC would be well below the RG 1.17 4 limit of 1 E-06/year CDF and 1 E-07/year LERF allowed for individual STI changes by the SFCP. The NRC staff concludes that exclusion of the impact of SOKC on internal flooding results is acceptable because it would have minor impact on STI evaluations.

-12 -In RAl-2 (Reference 6), NRC staff noted that F&O IE-C6-01 from the 2009 internal events PRA full-scope peer review, concerning exclusion of initiating event fault tree model failure modes, did not appear to be resolved for this application and requested justification that the cited modelling exclusions would not impact STI evaluations. The NRC staff reviewed the licensee's response to RAl-2 (Reference 2), and noted that it was still not clear whether all failure modes cited in the F&O that could impact STI evaluations had been incorporated into the modeling. Therefore, the NRC staff requested further justification in a second RAl-2.01 submitted to the licensee by letter dated April 12, 2016 (Reference 7). In response to RAI 2.01 by letter dated April 28, 2016 (Reference 3), the licensee stated that the development of the revised model considered failure modes cited in F&O IE-C6-01, where many of those failures were added to the model and others were excluded due to limited contribution. The licensee further provided an estimate of the overall risk contribution of the initiating event fault trees to the total plant CDF and LERF and stated that the overall risk contribution is very small. The NRC staff concludes that exclusion of some initiating event fault trees related to failures of valves, breakers, sensors, transmitters, and flow diversion paths is acceptable because it would have minor impact on STI evaluations. Based on the licensee's assessments using the currently applicable PRA standard and revision of RG 1.200, the NRC staff concludes that the level of PRA quality, combined with the evaluation and disposition of gaps, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177, Revision 1 (Reference 10). 3.1.4.2 Scope of the PRA The changes to the Administrative Controls section of the TSs will require the licensee to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10, Revision 1, to determine its potential impact on risk (i.e., CDF and LERF) from internal events, fires, seismic events, other external events, and shutdown conditions. In cases where a PRA of sufficient scope or quantitative risk models are unavailable, the licensee uses bounding analyses, or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero. The licensee has developed PRA models for at-power internal events, internal flooding, and fires, as well as a shutdown PRA model. In accordance with NEI 04-10, Revision 1, the licensee will use these models to perform quantitative evaluations to support the development of changes to surveillance frequencies in the SFCP. This is acceptable because the NRG-approved methodology in NEI 04-10, Revision 1, allows for more refined analysis to be performed to support changes to surveillance frequencies in the SFCP. In RAl-3 (Reference 6), the NRC staff noted that the LAR seemed to indicate that use of the fire PRA would be limited to qualitative analysis. In response to the RAI 3 (Reference 2), the licensee explained that if an SSC modelled in the fire PRA meets the modelling criteria of NEI 04-10, Revision 1, Step 8, then the quantitative evaluation described in Step 12 of the guidance will be completed to evaluate the individual and cumulative impacts on CDF/LERF and the change in CDF/LERF.

-13 -In RAl-4 (Reference 6), the NRG staff noted that the LAR submittal does not discuss how the SFGP will assess the risk of shutdown events when evaluating STls and requested a description of how shutdown events would be assessed. In response to RAl-4 (Reference 2), the license explained that there is a shutdown PRA for Waterford 3 that will be used in accordance with guidance in NEI 04-10, Revision 1, to address the contribution of shutdown events in STI evaluations for SSGs that are included in the shutdown PRA model. In LAR Attachment 2 the licensee stated that no Waterford 3 PRA model or applications associated with external hazards (e.g, seismic, high wind, external flooding) exists that could provide quantitative insights to support the STI effort. The licensee further explained that a Waterford 3 seismic PRA was conducted as a one-time review in response to the NRG Individual Plant Examination of External Events (IPEEE) Program, but that the scope was limited to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks. For those cases where a particular PRA model does not exist for a given hazard group, a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change. In PRA RAl-5 (Reference 6), the NRG staff requested that the licensee explain the process for updating the IPEEE evaluations to incorporate new information when performing the qualitative or bounding analyses. In its response to the RAl-5 (Reference 2), the licensee explained that the internal procedure for maintaining and updating the PRA ensures that needed improvements to external event evaluations (separate from the PRA) are also considered during periodic updates of the PRA models. The licensee explained that recent industry experience, new research, changes in industry standards, and other information are reviewed during the periodic update and will be used as applicable in updating the external event evaluations. This supports the licensee's conclusion that the IPEEE analysis of external hazards, updated to incorporate new information, can be used to support the SFGP. The NRG staff notes that this approach is in accordance with NEI 04-10, Revision 1. Based on the above, the NRG staff concludes that through the application of NRG-approved NEI 04-10, Revision 1, the licensee's evaluation methodology is sufficient to ensure the scope of the risk contribution of each surveillance frequency change is properly identified for evaluation and is consistent with Regulatory Position 2. 3.2 of RG 1.177, Revision 1 (Reference 10). 3.1.4.3 PRA Modeling The licensee's methodology includes the determination of whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out. The methodology adjusts the failure probability of the impacted SSCs, including any impacted CCF modes, based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy consistent with guidance contained in RG 1.200, Revision 2 (Reference 11 ), and by sensitivity studies identified in NEI 04-10, Revision 1. Thus, the NRG staff concludes that through the application of NRG-approved NEI 04-10, Revision 1, the Waterford 3 PRA modeling is sufficient to ensure an acceptable evaluation of

-14 -risk for the proposed changes in surveillance frequency, and is consistent with Regulatory Position 2. 3. 3 of RG 1.177, Revision 1. 3.1.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in PRAs may include a standby time-related contribution and a cyclic demand-related contribution. LAR Attachment 2, Section 3.2, "Identification of Key Assumptions," states that determination of standby failure rates is a key source of uncertainty and therefore sensitivity studies will be performed on standby failure rates for STI evaluations. The NRC staff notes that NEI 04-10, Revision 1, provides guidance on how to adjust the time-related failure contribution of SSCs and how to perform needed sensitivity studies (i.e., Step 14). NEI 04-10, Revision 1, criteria adjust the time-related failure contribution of SSCs affected by the proposed change to a surveillance frequency. This is consistent with RG 1.177, Revision 1, Section 2.3.3 (Reference 10), which permits separation of the failure rate contributions into demand and standby for evaluation of SRs. If the available data do not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The SSC failure rate (per unit time) is assumed to be unaffected by the change in test frequency, such that the failure probability is assumed to increase linearly with time, and will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The NEI 04-10 process requires consideration of qualitative sources of information with regards to potential impacts of test frequency on SSC performance, including industry and plant-specific operating experience, vendor recommendations, industry standards, and code-specified test intervals. Thus, the process is not reliant upon risk analyses as the sole basis for the proposed changes. The potential benefits of a reduced surveillance frequency, including reduced downtime and reduced potential for restoration errors, test-caused transients, and test-caused wear of equipment, are identified qualitatively, but not quantitatively assessed. Thus, the NRC staff concludes that through the application of NRC-approved NEI 04-10, Revision 1, the licensee has employed reasonable assumptions with regard to extensions of STls, and is consistent with Regulatory Position 2.3.4 of RG 1.177, Revision 1. NEI 04-10, Revision 1, requires performance monitoring of SSCs whose surveillance frequencies have been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of Maintenance Rule monitoring of equipment performance. In the event of SSC performance degradation, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may be required by the Maintenance Rule. 3.1.4.5 Sensitivity and Uncertainty Analyses By specifying that the TSs require that changes to the frequencies listed in the SFCP be made in accordance with NEI 04 10, Revision 1, the licensee will be required to conduct sensitivity studies that assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact on the frequency of initiating events, and any identified deviations from CC II of the PRA standard. Where the sensitivity analyses identify

-15 -a potential impact on the proposed change, revised surveillance frequencies are considered, along with any qualitative considerations that may bear on the results of such sensitivity studies. The licensee will also be required to perform monitoring and feedback of SSC performance once the revised surveillance frequencies are implemented. Thus, the NRC staff concludes that through the application of NRG-approved NEI 04-10, Revision 1, the licensee has appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations and is consistent with Regulatory Position 2.3.5 of RG 1.177, Revision 1. 3.1.4.6 Acceptance Guidelines The licensee will be required to quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using NEI 04-10, Revision 1, in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1 E-6 per year for change to CDF, and below 1 E-7 per year for change to LERF. These changes to CDF and LERF are consistent with the acceptance criteria of RG 1.17 4, Revision 2 (Reference 9), for very small changes in risk. Where the RG 1.17 4, Revision 2, acceptance criteria are not met, the process in NEI 04-10, Revision 1, either considers revised surveillance frequencies which are consistent with RG 1.17 4, Revision 2, or the process terminates without permitting the proposed changes. Where quantitative results are unavailable for comparison with the acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible or zero. Otherwise, bounding quantitative analyses are required which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.17 4, Revision 2 (Reference 9), acceptance guidelines for very small changes in risk. In addition to assessing each individual SSC surveillance frequency change, the cumulative impact of all changes must result in a risk impact less than 1 E-5 per year for change to CDF, and less than 1 E-6 per year for change to LERF, and the total CDF and total LERF must be reasonably shown to be less than 1 E-4 per year and 1 E-5 per year, respectively. These values are consistent with the acceptance criteria of RG 1.17 4, Revision 2, as referenced by RG 1 .177, Revision 1 (Reference 10), for changes to surveillance frequencies. Consistent with the NRC's SE dated September 19, 2007, for NEI 04-10, Revision 1, the TS SFCP will require the licensee to calculate the total change in risk (i.e., the cumulative risk) by comparing a baseline model that uses failure probabilities based on surveillance frequencies prior to being changed per the SFCP to a revised model that uses failure probabilities based on the changed surveillance frequencies. The NRC staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (i.e., less than 5E-8 per year for CDF and 5E-9 per year for LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies. The quantitative acceptance guidance of RG 1.17 4, Revision 2, is supplemented by qualitative information to evaluate the proposed changes to surveillance frequencies, including industry and plant-specific operating experience, vendor recommendations, industry standards, the results of sensitivity studies, and SSC performance data and test history. The final acceptability of the proposed change is based on all of these considerations and not solely on the PRA results. Post implementation performance monitoring and feedback are also required to assure

-16 -continued reliability of the components. The licensee's application of NRG-approved NEI 04-10, Revision 1, provides acceptable methods for evaluating the risk increase associated with proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177, Revision 1. Therefore, the NRC staff concludes that the proposed methodology satisfies the fourth key safety principle of RG 1.177, Revision 1, by assuring any increase in risk is small consistent with the intent of the Commission's Safety Goal Policy Statement. 3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's adoption of TSTF-425, Revision 3, requires application of NRG-approved NEI 04-10, Revision 1, in the SFCP. NEI 04-10, Revision 1, requires performance monitoring of SSCs whose surveillance frequencies have been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback include consideration of Maintenance Rule monitoring of equipment performance. In the event of SSC performance degradation, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may be required by the Maintenance Rule. The performance monitoring and feedback specified in NEI 04-10, Revision 1, is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 of RG 1.177, Revision 1. Thus, the NRC staff concludes that the fifth key safety principle of RG 1.177, Revision 1, is satisfied. 3.2 Addition of Surveillance Frequency Control Program to Administrative Controls The licensee proposed including the SFCP and specific requirements into the Waterford 3 TSs, Section 6.5.18, as follows: Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure that the associated Limiting Conditions for Operation are met. a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program. b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies," Revision 1. c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

-17 -The proposed program is consistent with the model application of TSTF-425, and therefore, the staff concludes that it is acceptable. 3.3 Deviations from TSTF-425 and Other Changes In Attachment 1, Section 2.2 of the LAR, the licensee identified optional changes and variations from the approved TSTF-425, Revision 3. The NRC staff reviewed the changes and variations and made the following determinations: 1. The insert provided in TSTF-425 pertaining to text that describes each frequency relocated to the SFCP has been revised from "The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program" to read "The Surveillance Frequency(ies) is/are based on operating experience, equipment reliability, and plant risk and is/are controlled under the Surveillance Frequency Control Program." As the licensee stated in paragraph 1 of Section 2.2 of the LAR, this variation was consistent with the inserts proposed by TSTF-425, Revision 3. The NRC staff determined that this is, therefore acceptable and the SRs will continue to meet 10 CFR 50.36(c)(3). 2. In the guidance in NUREG-1432, Revision 4, the Administrative Controls are located in Section 5.0. The approved programs for Waterford 3 are described in Section 6.0, "Administrative Controls," of the Technical Specifications. Therefore, the Surveillance Frequency Control Program description will be located at 6.5.18 of the Waterford 3 TS in lieu of at 5.5.20 as in the NUREG-1432 guidance. The description of the program is the same as in the approved TSTF-425, Revision 3. The NRC staff determined that this change is administrative in nature and is an acceptable variation from the approved TSTF tailored to fit the Waterford 3 TSs. Therefore the SRs will continue to meet 10 CFR 50.36(c)(3). 3. The licensee additionally proposed to remove a reference to the prescribed surveillance frequency ("at a frequency of 18 months on a STAGGERED TEST BASIS") of SR 4.7.6.1.b from TS 6.5.17 paragraph d. Removal of this language from paragraph d was necessary so that the frequency of SR 4.7.6.1.b could be controlled by the new surveillance frequency control program of TSTF-425, Revision 3 upon approval of the LAR. The NRC staff determined that removal of the language does not otherwise affect the readability or intent of TS 6.5.17; therefore this change is acceptable and consistent with TSTF-425, Revision 3. TS 6.5.17 will continue to meet 10 CFR 50.36(c)(3). 4. Attachment 7 of the LAR contained a cross-reference between NUREG-1432 surveillances included in TSTF-425, Revision 3 versus Waterford 3 surveillances. The cross reference served only as an aid in the NRC staff review and provided information as stated in paragraph 4 of section 2.2 of the LAR. 5. Paragraphs 5, 7 and the conclusion of Section 2.2 of the LAR identified that the Waterford 3 TSs are custom and contain plant-specific SRs not included in the approved TSTF-425, Revision 3. Approved TSTF-425, Revision 3 states, "The

-18 -proposed change relocates all periodic Surveillance Frequencies from the Technical Specifications and places the Frequencies under licensee control in accordance with a new program" and "All surveillances are relocated except. .. [four exclusion criteria for the surveillance frequencies are listed]." It does not add, delete, or modify the content of the surveillance requirements themselves. These statements denote that TSTF-425, Revision 3, applies to all surveillances, including the Waterford 3 plant-specific surveillances, that are periodic and do not meet one of the exclusion criteria. Application of the TSTF to Waterford 3's custom TS format is in accordance with the Commission's final policy statement on TSs, as published in the Federal Register on July 22, 1993 (58 FR 39132). The NRC staff reviewed the marked-up SRs in the LAR and the supplement submitted by letter dated March 3, 2016, to ensure that no surveillances were included that matched the exclusion criteria. The NRC staff determined that all marked-up surveillances included in the original LAR as supplemented, or corrected in the supplement, were included within the scope of approved TSTF-425, Revision 3. Therefore the SRs will continue to meet 10 CFR 50.36(c)(3). 6. The NRC staff reviewed only the changes proposed by the licensee for Waterford 3. Paragraph 6 of Section 2.2 of the LAR mentions that there are SRs contained in NUREG-1432, Revision 4, and in approved TSTF-425, Revision 3 that do not appear in Waterford 3 TSs. These are not within the scope of this plant specific Waterford 3 change and therefore were not reviewed. 3.4 Summary and Conclusions The NRC staff has reviewed the licensee's proposed relocation of some surveillance frequencies to a licensee-controlled document, and controlling changes to surveillance frequencies in accordance with a new program, the SFCP, which is identified in the Administrative Controls of the TSs. The NRC staff confirmed that this amendment does not relocate surveillance frequencies that reference other approved programs for the specific interval, are purely event-driven, are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs, or are related to specific conditions. The SFCP and TSs Section 6.0, Subsection 6.5.18 references NEI 04-10, Revision 1, which provides a risk-informed methodology using plant-specific risk insights and performance data to revise surveillance frequencies within the SFCP. This methodology supports relocating surveillance frequencies from TSs to a licensee-controlled document, provided those frequencies are changed in accordance with the NRG-approved NEI 04-10, Revision 1, which is specified in the Administrative Controls of the TSs. The licensee-proposed adoption of TSTF-425, Revision 3, and the risk-informed methodology of NRG-approved NEI 04-10, Revision 1, as referenced in the Administrative Controls of TSs, satisfies the key principles of risk-informed decisionmaking applied to changes to TSs, as delineated in RG 1.177 and RG 1.17 4, in that:

The proposed change meets current regulations;

The proposed change is consistent with defense-in-depth philosophy;

-19 -* The proposed change maintains sufficient safety margins;

Increases in risk resulting from the proposed change are small and consistent with the Commission's Safety Goal Policy Statement; and

The impact of the proposed change is monitored with performance measurement strategies. Section 50.36(c) of 10 CFR discusses the categories that will be included in TSs. Section 50.36(c)(3) of 10 CFR discusses the specific category of SRs and states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The NRC staff finds that with the proposed relocation of surveillance frequencies to a licensee-controlled document and administratively controlled in accordance with the TS SFCP, the licensee continues to meet the requirements in 10 CFR 50.36.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes the surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on September 1, 2015 (80 FR 52805). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

-20 -7.0 REFERENCES 1. Chisum, M. R., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a License Controlled Program, Waterford Steam Electric Station, Unit 3, Docket No. 50-382, License No. NPF-38," dated June 17, 2015 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML 15170A121). 2. Chisum, M. R., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Responses to Request for Additional Information Regarding the Risk-Informed Surveillance Requirements License Amendment Request (LAR), Waterford Steam Electric Station, Unit 3 (Waterford 3), Docket No. 50-382, License No. NPF-38," dated March 3, 2016 (ADAMS Accession No. ML 16063A532). 3. Chisum, M. R., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Responses to Request for Additional Information Regarding the Risk-Informed Surveillance Requirements License Amendment Request (LAR), Waterford Steam Electric Station, Unit 3 (Waterford 3), Docket No. 50-382, License No. NPF-38," dated April 28, 2016 (ADAMS Accession No. ML 16119A428). 4. Nuclear Energy Institute, NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession No. ML071360456). 5. Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," dated March 18, 2009 (ADAMS Package Accession No. ML090850642). 6. Pulvirenti, A. L., U.S. Nuclear Regulatory Commission, letter to Site Vice President, Operations, Entergy Operations, Inc., "Waterford Steam Electric Station, Unit 3 -Request for Additional Information Regarding the Risk-Informed Surveillance Requirements License Amendment Request (CAC No. MF6366)," dated January 22, 2016 (ADAMS Accession No. ML 16015A294). 7. Pulvirenti, A. L., U.S. Nuclear Regulatory Commission, letter to Site Vice President, Operations, Entergy Operations, Inc., "Waterford Steam Electric Station, Unit 3 -Request for Additional Information Regarding the Risk-Informed Surveillance Requirements License Amendment Request (CAC No. MF6366)," dated April 12, 2016 (ADAMS Accession No. ML 16102A152). 8. Nieh, H. K., U.S. Nuclear Regulatory Commission, letter to B. Bradley, Nuclear Energy Institute, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, Risk-informed Technical Specification Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies (TAC No. MD6111 ),"dated September 19, 2007 (ADAMS Accession No. ML072570267).

-21 -9. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Specific Changes to the Licensing Basis," May 2001 (ADAMS Accession No. ML 100910006). 10. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.177, "An Approach for Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 1, May 2011 (ADAMS Accession No. ML 100910008). 11. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Informed Activities," Revision 2, March 2009 (ADAMS Accession No. ML090410014). 12. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant Specific Changes to the Licensing Basis: General Guidance," June 2007 (ADAMS Accession No. ML071700658). 13. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Chapter 19, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests after Initial Fuel Load," Revision 3, September 2012 (ADAMS Accession No. ML 12193A107). 14. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Chapter 16, Section 16.1, "Risk-Informed Decision Making: Technical Specifications," Revision 1, March 2007 (ADAMS Accession No. ML070380228). 15. American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) RA-Sa-2009, "Addenda to ASME RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2, 2009, New York, NY. 16. Nuclear Energy Institute, NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision 1, May 2006 and NEI 00-02 Appendix D, "Self Assessment Process for addressing ASME PRA Standard RA-Sb-2005, as endorsed by NRC Regulatory Guide 1.200," October 2006 (ADAMS Accession Nos. ML061510619 and ML063390593, respectively). 17. Nuclear Energy Institute, NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 2, November 2008 (ADAMS Accession No. ML083430462). 18. American Society of Mechanical Engineers (ASME) PRA Standard ASME RA-Sb-2005, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," New York, NY.

-22 -19. Jacobs, D., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2201 Edition), Waterford Steam Electric Station, Unit 3, Docket No. 50-382, License No. NPF-38," dated November 17, 2011 (ADAMS Accession No. ML 113220230). 20. Chisum, M. R., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Responses to Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA 805 License Amendment Request (LAR) Waterford Steam Electric Station, Unit 3 (Waterford 3), Docket No. 50-382, License No. NPF-38," dated May 14, 2015 (ADAMS Accession No. ML 15138A057). 21. Chisum, M. R., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Supplement to the Risk-Informed Surveillance Requirements License Amendment Request (LAR) Waterford Steam Electric Station, Unit 3 (Waterford 3), Docket No. 50-382, License No. NPF-38," dated July 12, 2016 (ADAMS Accession No. ML 16194A452). Principal Contributors: M. Reisifard P. Snyder Date: July 26, 2016 July 26, 2016 Site Vice President Entergy Operations, Inc. Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3-ISSUANCE OF AMENDMENT RE: ADOPTION OF TSTF-425, REVISION 3 "RELOCATE SURVEILLANCE FREQUENCIES TO LICENSEE CONTROL -RITSTF INITIATIVE 5b" (CAC NO. MF6366)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 249 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated June 17, 2015, as supplemented by letters dated March 3, April 28, and July 12, 2016. The amendment modifies the TSs by relocating specific surveillance frequencies to a licensee-controlled program. The proposed changes are consistent with the NRG-approved Technical Specifications Task Force (TSTF) Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF [Risk-Informed TSTF] Initiative 5b." A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Docket No. 50-382

Enclosures:

Sincerely, IRA/ April L. Pulvirenti, Project Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 1. Amendment No. 249 to NPF-38 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: PUBLIC LPL4-2 R/F RidsAcrs_MailCTR Resource RidsNrrDeEeeb Resource RidsNrrLAPBlechman Resource RidsNrrDorlLPL4-2 Resource RidsNrrDraApla Resource RidsNrrDssStsb Resource RidsNrrDorlDpr Resource NRRADAMSA ccess1on N ML161 9A 19 o. 5 4 OFFICE NRR/DORL/LPL4-2/PM NRR/DORL/LPL/4-2/LA NAME APulvirenti PBlechman DATE 06/09/2016 06/9/2016 OFFICE NRR/DE/EEEB NRR/DE/EEEB RidsNrrPMWaterford Resource RidsRgn4MailCenter Resource MReisifard, NRR TMarinez-Navedo, NRR *b 1y memo d t d ae NRR/DRA/APLA/BC* SRosenberg 5/26/2016 NRR/DE/EEEB NAME SRav (non-concur) TMartinez-Navedo (non-concur) NOtto (non-concur) DATE 7/7/16 7/7/16 7/7/16 OFFICE NRR/DE/EEEB NRR/DE/EEEB NRR/DE/EEEB/BC NAME SSom (non-concur) RFitzpatrick (non-concur) JZimmerman (non-concur) DATE 7/7/16 7/7/16 7/14/16 OFFICE NRR/DORL/LPL4-2/LA NRR/DORL/LPL4-2/BC(A) NRR/DOR/LPL4-2/PM NAME PBlechman SAnderson APulvirenti DATE 07/18/2016 07/25/2016 7/26/16 OFFICIAL RECORD COPY PSnyder, NRR GMatharu, NRR RMathew, NRR SRay, NRR **b *1 d t d 1yema1 ae NRR/DSS/STSB/BC* AKlein 07/20/2016 NRR/DE/EEEB RMathew (non-concur) 7/7/16 OGC --NLO JLindell 06/17/2016

v t e Letter Sequence Approval
CAC:MF6366, (Approved, Closed)
Initiation Acceptance... Supplement Administration Questions 22 January 2016 12 April 2016 Answers 3 March 2016 28 April 2016 Results Approval Other: ML16033A197, ML16209A045
MONTHYEAR2016-04-12 [Table View]

ML16159A419

Text

SUBJECT:

Dear Sir or Madam:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

SUBJECT:

Dear Sir or Madam:

Enclosures:

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