ML15280A242

Issuance of Amendment No. 324 Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, Adoption of TSTF-425,Rev 3

ML15280A242
Person / Time
Site:	Millstone
Issue date:	10/29/2015
From:	Richard Guzman Plant Licensing Branch 1
To:	Heacock D Dominion Nuclear
Guzman R
References
TAC MF5096
Download: ML15280A242 (125)

Text

{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 29, 2015 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM, ADOPTION OF TSTF-425, REVISION 3 (TAC NO. MF5096)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 324 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated October 22, 2014, as supplemented by letters dated June 5, July 20, and August 27, 2015. The amendment modifies the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee controlled program with the adoption of Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - [Risk-Informed Technical Specification Task Force (RITSTF)] Initiative Sb." Additionally, the amendment adds a new program, the Surveillance Frequency Control Program, to TS Section 6, Administrative Controls. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Richard V. Guzman, Sr. Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336

Enclosures:

1. Amendment No. 324 to DPR-65

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT, INC. DOCKET NO. 50-336 MILLSTONE POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 324 Renewed License No. DPR-65

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

A. The application for amendment by Dominion Nuclear Connecticut, Inc. (the licensee) dated October 22, 2014, as supplemented on June 5, July 20, and August 27, 2015, 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 Renewed Facility Operating License No. DPR-65 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 324, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance, and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: October 29, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 324 RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. Remove Insert 3 3 Replace the following pages of the 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. Remove Insert 1-4 1-4 1-9 1-9 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-3 3/4 1-7 3/4 1-7 3/4 1-21 3/4 1-21 3/4 1-25 3/4 1-25 3/4 1-26 3/4 1-26 3/4 1-27 3/4 1-27 3/4 1-29 3/4 1-29 3/4 1-31 3/4 1-31 3/4 2-2 3/4 2-2 3/4 2-9 3/4 2-9 3/42-10 3/4 2-10 3/4 2-13 3/4 2-13 3/4 3-1 3/4 3-1 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 3/4 3-20 3/4 3-20 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-35 3/4 3-35

Remove Insert 3/4 4-1 3/4 4-1 3/4 4-1a 3/4 4-1a 3/4 4-1 c 3/4 4-1c 3/4 4-1e 3/4 4-1 e 3/4 4-1g 3/4 4-1g 3/4 4-1 h 3/4 4-1 h 3/4 4-3a 3/4 4-3a 3/4 4-4 3/4 4-4 3/4 4-8a 3/4 4-8a 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 3/4 4-14 3/4 4-14 3/4 4-18 3/4 4-18 3/4 4-21 b 3/4 4-21 b 3/4 5-2 3/4 5-2 3/4 5-4 3/4 5-4 3/4 5-5 3/4 5-5 3/4 5-8 3/4 5-8 3/4 5-9 3/4 5-9 3/4 6-1 3/4 6-1 3/4 6-6a 3/4 6-6a 3/4 6-8 3/4 6-8 3/4 6-9 3/4 6-9 3/4 6-12 3/4 6-12 3/4 6-13 3/4 6-13 3/4 6-15 3/4 6-15 3/4 6-19 3/4 6-19 3/4 6-24 3/4 6-24 3/4 6-25 3/4 6-25 3/4 6-26 3/4 6-26 3/4 6-27 3/4 6-27 3/4 6-28 3/4 6-28 3/4 7-5 3/4 7-5 3/4 7-5a 3/4 7-5a 3/4 7-6 3/4 7-6 3/4 7-7 3/4 7-7 3/4 7-8 3/4 7-8 3/4 7-9b 3/4 7-9b 3/4 7-9c 3/4 7-9c 3/4 7-9d 3/4 7-9d 3/4 7-11 3/4 7-11 3/4 7-12 3/4 7-12 3/4 7-17 3/4 7-17 3/4 7-17a 3/4 7-17a 3/4 7-34 3/4 7-34 3/4 8-2a 3/4 8-2a 3/4 8-3 3/4 8-3

Remove Insert 3/4 8-3a 3/4 8-3a 3/4 8-4 3/4 8-4 3/4 8-6 3/4 8-6 3/4 8-6a 3/4 8-6a 3/4 8-7 3/4 8-7 3/4 8-8 3/4 8-8 3/4 8-9 3/4 8-9 3/4 8-10 3/4 8-10 3/4 8-11 3/4 8-11 3/4 9-1 3/4 9-1 3/4 9-2 3/4 9-2 3/4 9-5 3/4 9-5 3/4 9-8a 3/4 9-8a 3/4 9-8c 3/4 9-8c 3/4 9-11 3/4 9-11 3/4 9-12 3/4 9-12 3/4 9-19 3/4 9-19 3/4 9-21 3/4 9-21 3/4 10-2 3/4 10-2 6-33 6-33 6-34

Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and 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 The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 324, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications. Renewed License No. DPR-65 Amendment No. 324

DEFINITIONS AZIMUTHAL POWER TILT - T q 1.18 AZIMUTHAL POWER TILT shall be the difference between the maximum power generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core. AZIMUTHALPOWERTILT = [Maximum power in any core quadrant (upper or lower)]- I A verage power of all quadrants (upper or lower) DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration ofl-131 (micro-curie/gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, I-132, I-133, I-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion." DOSE EQUIVALENT XE-133 1.20 DOSE EQUIVALENT XE-133 shall be that concentration ofXe-133 (micro-curie/gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-13 lm, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.l of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil." 1.21 Deleted FREQUENCY NOTATION 1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. MILLSTONE - UNIT 2 1-4 Amendment No. -W4, m, m, 3-{f'.7, 324

TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY s At least once per 12 hours. D At least once per 24 hours. w At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. SA At least once per 6 months. R At least once per 18 months. SIU Prior to each reactor startup. p Prior to each release. N.A. Not applicable. SFCP At the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 1-9 Amendment No. -l-Q.4, 324

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1.1 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM) w LIMITING CONDITION FOR OPERATION 3.1. l. l The SHUTDOWN MARGIN shall be within the limit specified in the CORE OPERATING LIMITS REPORT. APPLICABILITY: MODES 3(!)*, 4 and 5. ACTION: With the SHUTDOWN MARGIN not within the limit specified in the CORE OPERATING LIMITS REPORT, within 15 minutes, initiate and continue boration at~ 40 gpm of boric acid solution at or greater than the required refueling water storage tank (RWST) concentration (ppm) until the SHUTDOWN MARGIN is restored to within limit. SURVEILLANCE REQUIREMENTS 4.1.1.1 Verify SHUTDOWN MARGIN is within the limit specified in the CORE OPERATING LIMITS REPORT at the frequency specified in the Surveillance Frequency Control Program.

• (l)See Special Test Exception 3.10.1 MILLSTONE UNIT 2w Amendment No.~. 6+, ::fl:, .+4, -89,

                                                                                        -l48,~,324

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 REACTIVITY CONTROL SYSTEMS REACTIVITY BALANCE LIMITING CONDITION FOR OPERATION 3.1. l.2 The core reactivity balance shall be within +/- 1% &!k of predicted values. APPLICABILITY: MODES 1 and 2. ACTION: With core reactivity balance not within limit: Re-evaluate core design and safety analysis and determine that the reactor core is acceptable for continued operation and establish appropriate operating restrictions and Surveillance Requirements within 7 days or otherwise be in MODE 3 within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.1.1.2 Verify*(I) overall core reactivity balance is within+/- 1% dk/k of predicted values prior to entering MODE 1 after fuel loading and at the frequency specified in the Surveillance Frequency Control Program**(2). The provisions of Specification 4.0.4 are not applicable.

• (1) The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel bumup of 60 Effective Full Power Days after each fuel loading.
  • (2) Only required after 60 Effective Full Power Days.

MILLSTONE - UNIT 2 3/4 1-3 Amendment No. 48, ~' 324

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION

3. l.1.5 The Reactor Coolant System temperature (Tavg) shall be~ 5 l 5°F when the reactor is critical.

APPLICABILITY: MODES 1 and 2 *. ACTION: With the Reactor Coolant System temperature (Tavg) < 515°F, restore Tavg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.1.5 The Reactor Coolant System temperature (Tavg) shall be determined to be ~ 5 l 5°F.

a. Within 15 minutes prior to making the reactor critical, and

b. At the frequency specified in the Surveillance Frequency Control Program when the reactor is critical and the Reactor Coolant System temperature (Tavg) is<

525°F.

• With Keff~ 1.0.

MILLSTONE - UNIT 2 314 1-7 Amendment No. ~. ~. 324

REACTIVITY CONTROL SYSTEMS ACTION: (Continued): C. CEA Deviation Circuit C.1 Verify the indicated position of each CEA to be within inoperable. 10 steps of all other CEAs in its group within 1 hour and every 4 hours thereafter or otherwise be in MODE 3 within the next 6 hours. D. One or more CEAs untrippable. D.1 Be in MODE 3 within 6 hours. OR Two or more CEAs misaligned by

~ 20 steps.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 Verify the indicated position of each CEA to be within 10 steps of all other CEAs in its group at the frequency specified in the Surveillance Frequency Control Program AND within 1 hour following any CEA movement larger than 10 steps. 4.1.3.1.2 Verify CEA freedom of movement (trippability) by moving each individual CEA that is not fully inserted into the reactor core 10 steps in either direction at the frequency specified in the Surveillance Frequency Control Program. 4.1.3.l.3 Verify the CEA Deviation Circuit is OPERABLE at the frequency specified in the Surveillance Frequency Control Program by a functional test of the CEA group Deviation Circuit which verifies that the circuit prevents any CEA from being misaligned from all other CEAs in its group by more than 10 steps (indicated position). 4.1.3.1.4 Verify the CEA Motion Inhibit is OPERABLE by a functional test which verifies that the circuit maintains the CEA group overlap and sequencing requirements of Specification 3.1.3.6 and that the circuit prevents regulating CEAs from being inserted beyond the Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT:

a. Prior to each entry into MODE 2 from MODE 3, except that such verification need not be performed more often than once per 31 days, and

b. At the frequency specified in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 314 1-21 Amendment No. ~ ~, 324

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS (Continued) LIMITING CONDITION FOR OPERATION (Continued) b) The CEA group(s) with the inoperable indicator is fully inserted, and subsequently maintained fully inserted, while maintaining the withdrawal sequence and THERMAL POWER level required by Specification 3.1.3.6 and when this CEA group reaches its fully inserted position, the "Full In" limit of the CEA with the inoperable position indicator is actuated and verifies this CEA to be fully inserted. Subsequent operation shall be within the limits of Specification 3.1.3.6.

4. If the failure of the position indicator channel(s) is during STARTUP, the CEA group(s) with the inoperable position indicator channel must be moved to the "Full Out" position and verified to be fully withdrawn via a "Full Out" indicator within 4 hours.

c. With a maximum of one reed switch position indicator channel per group or one pulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel at either its fully inserted position or fully withdrawn position, operation may continue provided:

1. The position of this CEA is verified immediately and at least once per 12 hours thereafter by its "Full In" or "Full Out" limit (as applicable).

2. The fully inserted CEA group(s) containing the inoperable position channel is subsequently maintained fully inserted, and

3. Subsequent operation is within the limits of Specification 3.1.3.6.

d. With one or more pulse counting position indicator channels inoperable, operation in MODES 1 and 2 may continue for up to 24 hours provided all of the reed switch position indicator channels are OPERABLE.

SURVEILLANCE REQUIREMENTS 4.1.3.3 Each required position indicator channel shall be determined to be OPERABLE by verifying the pulse counting position indicator channels and the reed switch position indicator channels agree within 6 steps at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 1-25 Amendment No. -H+, ~' 324

REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual CEA drop time, from a fully withdrawn position, shall be 5 2.75 seconds from when electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a. Tavg ~ 515°F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2. ACTION: With the drop time of any CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2. SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time shall be demonstrated through measurement with Tavg ~ 515°F, and all reactor coolant pumps operating prior to reactor criticality:

a. For all CEAs following each removal of the reactor vessel head,

b. For specifically affected individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At the frequency specified in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 3/4 1-26 Amendment No. 38-, £, 99, m, ~. 324

REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to~ 176 steps. APPLICABILITY: MODE 1*(!) MODE 20),(2)** with any regulating CEA not fully inserted. ACTION: INOPERABLE EQUIPMENT REQUIRED ACTION A. One or more shutdown CEAs not A.1 Restore shutdown CEA(s) to within limit. within limit within 2 hours or otherwise be in MODE 3 within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3.5 Verify each shutdown CEA is withdrawn~ 176 steps at the frequency specified in the Surveillance Frequency Control Program.

• (1) This LCO is not applicable while performing Specification 4.1.3.1.2.
  • (2)See Special Test Exceptions 3.10.1 and 3.10.2.

MILLSTONE - UNIT 2 314 1-27 Amendment No.~. 324

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS (Continued) B. Regulating CEA groups B.l Verify Short Term Steady State Insertion Limits as inserted between the Long Tenn specified in the CORE OPERATING LIMITS REPORT Steady State Insertion limit and are not exceeded within 15 minutes or otherwise be in the Transient Insertion Limit MODE 3 within the next 6 hours. specified in the CORE OPERATING LIMITS REPORT for intervals > 4 hours per 24 hour interval. B.2 Restrict increases in THERMAL POWER to < 5% RATED THERMAL POWER per hour within 15 minutes or otherwise be in MODE 3 within the next 6 hours. C. Regulating CEA groups C. l Restore regulating CEA groups to within the Long inserted between the Long Term Term Steady State Insertion Limit specified in the CORE Steady State Insertion Limit and OPERATING LIMITS REPORT within 2 hours or the Transient Insertion Limit otherwise be in MODE 3 within the next 6 hours. specified in the CORE OPERATING LIMITS REPORT for intervals > 5 effective full power days (EFPD) per 30 EFPD or interval> 14 EFPD per 365 EFPD. D. PDIL alarm circuit D.l Perform Specification 4.1.3.6.l within 1 hour and inoperable. once per 4 hours thereafter or otherwise be in MODE 3 within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3 .6.1 Verify each regulating CEA group position is within the Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT at the frequency specified in the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable for entering into MODE 2 from MODE 3. 4.1.3.6.2 Verify the accumulated times during which the regulating CEA groups are inserted beyond the Steady State Insertion Limits but within the Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT at the frequency specified in the Surveillance Frequency Control Program. 4.1.3.6.3 Verify PDIL alarm circuit is OPERABLE at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 1-29 Amendment No. -+/-48, £, ~. ~' 324

REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized. APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentration is less than refueling concentration of Specification 3. 9 .1. ACTION: With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours or immediately open the reactor trip circuit breakers. SURVEILLANCE REQUIREMENTS 4.1.3. 7 The control rod drive mechanisms shall be verified to be de-energized at the frequency specified in the Surveillance Frequency Control Program.

• The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 500° F, the pressurizer pressure is greater than 2000 psia and the requirements of Limiting Condition for Operation for Specification 3.3.1.1, "Reactor Protective Instrumentation," are met.

MILLSTONE - UNIT 2 3/4 1-31 Amendment No. He, 291, 3++, 324

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.1.2 Excore Detector Monitoring System *( 1) - The excore detector monitoring system may be used for monitoring the core power distribution by:

a. Verifying at the frequency specified in the Surveillance Frequency Control Program that the CEAs are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3. 6.

b. Verifying at the frequency specified in the Surveillance Frequency Control Program that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the allowable limits specified in the CORE OPERATING LIMITS REPORT.

4.2.1.3 Incore Detector Monitoring System**<2), ***< 3) - The incore detector monitoring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a. Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at the frequency specified in the Surveillance Frequency Control Program.

b. Have their alarm setpoint adjusted to less than or equal to the limits specified in the CORE OPERATING LIMITS REPORT at the frequency specified in the Surveillance Frequency Control Program.

• (l) Only required to be met when the Excore Detector Monitoring System is being used to determine Linear Heat Rate.
  • ( 2)0nly required to be met when the Incore Detector Monitoring System is being used to determine Linear Heat Rate.
    • (3)Not required to be performed below 20% RATED THERMAL POWER.

MILLSTONE - UNIT 2 3/4 2-2 Amendment No. ti,~.£, 99, H-9,

                                                                                               -14&, +/-8G, 324

POWER DISTRIBUTION LIMITS TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR- FTr LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of FTr shall be within the 100% power limit specified in the CORE OPERATING LIMITS REPORT. The F Tr value shall include the effect of AZIMUTHAL POWER TILT. APPLICABILITY: MODE 1 with THERMAL POWER >20% RTP*. ACTION: With FTr exceeding the 100% power limit within 6 hours either:

a. Reduce THERMAL POWER to bring the combination of THERMAL POWER and FTr to within the power dependent limit specified in the CORE OPERATING LIMITS REPORT and withdraw the CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or

b. Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 F\ shall be determined to be within the l 00% power limit at the following intervals:

a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,

b. At the frequency specified in the Surveillance Frequency Control Program in MODE 1, and

c. Within four hours if the AZIMUTHAL POWER TILT (Tq) is> 0.020.

4.2.3.3 FTr shall be determined by using the incore detectors to obtain a power distribution map with all CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump Combination.

• See Special Test Exception 3.10.2.

MILLSTONE - UNIT 2 3/4 2-9 Amendment No. 3-8-, ~. :::µ:), 00, 99, ill,B-9,!48,!M'.,M4,~,~.~,324

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - Ta LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (Tq) shall be~ 0.02. APPLICABILITY: MODE 1 with THERMAL POWER> 50% of RATED THERMAL POWER(I)*. ACTION:

a. With the indicated T q > 0.02 but::::;; 0.10, either restore Tg_ to ::; 0.02 within 2 hours or verify the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours and once per 8 hours thereafter. Or otherwise, reduce THERMAL POWER to ::; 50% of RATED THERMAL POWER within the next 4 hours.

b. With the indicated Tq > 0.10, perform the following actions: (2)**

l. Verify the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR (FTr) is within the limit of Specification 3.2.3 within 2 hours; and

2. Reduce THERMAL POWER to~ 50% of RATED THERMAL POWER within 2 hours; and

3. Restore Tq::; 0.02 prior to increasing THERMAL POWER. Correct the cause of tb.e out oflimit condition prior to increasing THERMAL POWER.

Subsequent power operation above 50% of RATED THERMAL POWER may proceed provided that the measured T is verified::; 0.02 at least once 9 per hour for 12 hours, or until verified at 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.4.1 Verify Tq is within limit at the frequency specified in the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable for entering into MODE 1 with THERMAL POWER> 50% of RATED THERMAL POWER from MODE 1.

• ( 1)See Special Test Exception 3.10.2.
• *(2)All subsequent Required ACTIONS must be completed if power reduction commences prior to restoring Tq::; 0.10.

MILLSTONE - UNIT 2 314 2-10 Amendment No. 3%, £, 9G, H-9,+M-,

                                                                                                   ~.m,324

POWER DISTRIBUTION LIMITS DNBMARGIN LIMITING CONDITION FOR OPERATION 3.2.6 The DNB margin shall be preserved by maintaining the cold leg temperature, pressurizer pressure, reactor coolant flow rate, and AXIAL SHAPE INDEX within the limits specified in the CORE OPERATING LIMITS REPORT. APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its specified limits, restore the parameter to within its above specified limits within 2 hours or reduce THERMAL POWER to ~ 5% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.6.1 The cold leg temperature, pressurizer pressure, and AXIAL SHAPE INDEX shall be determined to be within the limits specified in the CORE OPERATING LIMITS REPORT at the frequency specified in the Surveillance Frequency Control Program. The reactor coolant flow rate shall be determined to be within the limit specified in the CORE OPERATING LIMITS REPORT at the frequency specified in the Surveillance Frequency Control Program. 4.2.6.2 The provisions of Specification 4.0.4 are not applicable. MILLSTONE - UNIT 2 3/4 2-13 Amendment No. 3%, ~. +H, -148, 324

3/4.3 INSTRUMENTATION 3/4.3. l REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3 .3 .1.1 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 4.3.1.1.1 Each required reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1. 4.3 .1.1.2 The bypass function and automatic bypass removal function shall be demonstrated OPERABLE during a CHANNEL FUNCTIONAL TEST once within 92 days prior to each reactor startup. The total bypass function shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3 .1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at the frequency specified in the Surveillance Frequency Control Program. Neutron detector$ are exempt from response time testing. Each test shall include at least one channel per function. MILLSTONE - UNIT 2 3/4 3-1 Amendment No. =R:, 98, m, 3-9+, 324

e; TABLE 4.3-1 ~ CZl REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 ~ I CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE fi ....... FUNCTIONAL UNIT CHECK CALIBRATION TEST ~UIRED N 1. Manual Reactor Trip N.A. N.A. S/U(l) N.A.

2. Power Level - High

a. Nuclear Power SFCP SFCP(2), SFCP 1, 2, 3*

SFCP(3 ),SFCP(5) w ~ b. ~T Power SFCP SFCP(4), SFCP SFCP 1 wI

3. Reactor Coolant Flow - Low SFCP SFCP SFCP 1, 2

°'

4. Pressurizer Pressure - High SFCP SFCP SFCP 1, 2

5. Containment Pressure - High SFCP SFCP SFCP 1, 2

6. Steam Generator Pressure - Low SFCP SFCP SFCP 1, 2

7. Steam Generator Water SFCP SFCP SFCP 1, 2 Level- Low

8. Local Power Density - High SFCP SFCP SFCP 1 8

(1) 9. Thermal Margin/Low Pressure SFCP SFCP SFCP 1, 2 t:::l 0.. s (1)

10. Loss ofTurbine--Hydraulic N.A. SFCP S/U(l) N.A.

!:I. Fluid Pressure - Low z~ u~ u~ ~v.; N

s::...... TABLE 4.3-1 (Continued) t:""" t:""" (/) REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS ....:i

~

tr! CHANNEL MODES IN WHICH I CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

~......            FUNCTIONAL UNIT           CHECK      CALIBRATION             TEST          REQUIRED
 ....:i N       11. Wide Range Logarithmic Neutron   SFCP          SFCP(5)            S/U(l)            3,4,5 Flux Monitor - Shutdown

12. DELETED

13. Reactor Protection System N.A. N.A. SFCP and S/U(l) I, 2 and*

VJ Logic Matrices ~ VJ -..J I 14. Reactor Protection System N.A. N.A. SFCP and S/U(l) 1, 2 and* Logic Matrix Relays

15. Reactor Trip Breakers N.A. N.A. SFCP 1, 2 and*

~

(1) t::S sm. z~ ~a; Jt ~~ ~w N

  .i:..

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system 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 engineered safety feature actuation system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours or declare the channel inoperable and take the ACTION shown in Table 3.3-3.

b. With an engineered safety feature actuation system instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each required engineered safety feature actuation system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-2. 4.3 .2.1.2 The bypass function and automatic bypass removal function shall be demonstrated OPERABLE during a CHANNEL FUNCTIONAL TEST onc.e within 92 days prior to each reactor startup. The total bypass function shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by bypass operation. MILLSTONE - UNIT 2 3/4 3-9 Amendment No. +98-, ~. ~. 3-e+, 324

INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function shall be demonstrated to be within the limit at the frequency specified in the Surveillance Frequency Control Program. Each test shall include at least one channel per function. MILLSTONE - UNIT 2 3/4 3-10 Amendment No. 49, ~. ~' m, 324

TABLE4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS -~ t""' ~ FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE REQUIRED 0 1. SAFETY INJECTION (SIAS)

~           a. Manual (Trip Buttons)         N.A.           N.A.        SFCP         N.A.

I

b. Containment Pressure - High SFCP SFCP SFCP 1, 2, 3

~

N 2. c. d. Pressurizer Pressure - Low Automatic Actuation Logic CONTAINMENT SPRAY (CSAS) SFCP N.A. SFCP N.A. SFCP SFCP(l) 1, 2, 3 1, 2, 3

a. Manual (Trip Buttons) N.A. N.A. SFCP N.A.

b. Containment Pressure-- SFCP SFCP SFCP 1, 2, 3 High- High VJ c. Automatic Actuation Logic N.A. N.A. SFCP(l) 1, 2, 3

~ 3. CONTAINMENT ISOLATION VJ N I (CIAS) 0

a. Manual CIAS (Trip Buttons) N.A. N.A. SFCP N.A.

b. Manual SIAS (Trip Buttons) N.A. N.A. SFCP N.A.

c. Containment Pressure - High SFCP SFCP SFCP 1, 2, 3

d. Pressurizer Pressure - Low SFCP SFCP SFCP 1, 2, 3

e. Automatic Actuation Logic N.A. N.A. SFCP(l) 1, 2, 3

4. MAIN STEAM LINE ISOLATION

a. Manual (Trip Buttons) N.A. N.A. SFCP N.A.

a b. Containment Pressure - High SFCP SFCP SFCP 1, 2, 3

~

i

c. Steam Generator Pressure - SFCP SFCP SFCP 1, 2, 3 0.. Low a

g d. Automatic Actuation Logic N.A. N.A. SFCP(l) 1, 2, 3 z~ 5. ENCLOSURE BUILDING FILTRATION (EBFAS) J~ a. b. Manual EBFAS (Trip Buttons) Manual SIAS (Trip Buttons) N.A. N.A. N.A. N.A. SFCP SFCP N.A. N.A. v~ c. Containment Pressure - High SFCP SFCP SFCP 1, 2, 3

d. Pressurizer Pressure - Low SFCP SFCP SFCP 1, 2, 3

~ \.>.)

e. Automatic Actuation Logic N.A. N.A. SFCP(l) 1, 2, 3 N

s;:: TABLE 4.3-2 (Continued) l' l' r:/l

      ....j ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0

ztrJ CHANNEL MODES IN WHICH I CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

        ~
        ........ FUNCTIONAL UNIT                         CHECK      CALIBRATION         TEST      ___MQUIRED
        ....j N        6. CONTAINMENT SUMP RECIRCULATION (SRAS)

a. Manual SRAS (Trip Buttons) N.A. N.A. SFCP N.A.

b. Refueling Water Storage SFCP SFCP SFCP 1, 2, 3 w Tank-Low

     ~

VJ I c. Automatic Actuation Logic N.A. N.A. SFCP(l) 1, 2, 3 N

7. DELETED

8. LOSS OF POWER

a. 4.16 kv Emergency Bus SFCP SFCP SFCP 1, 2, 3 Undervoltage - level one

        ~

0

b. 4.16 kv Emergency Bus Undervoltage - level two SFCP SFCP SFCP 1, 2, 3

[ 80 9. AUXILIARY FEEDWATER a z 0

a. Manual N.A. N.A. SFCP N.A.

b. Steam Generator Level - Low SFCP SFCP SFCP 1, 2, 3 u~

c. Automatic Actuation Logic. N.A. N.A. SFCP 1, 2, 3 jti u~

10. STEAM GENERATOR BLOWDOWN u~ a. Steam Generator Level - Low SFCP SFCP SFCP 1, 2, 3

~* \.;.) N ~

TABLE 4.3-2 (Continued) TABLE NOTATION (1) The coincident logic circuits shall be tested automatically or manually at the frequency specified in the Surveillance Frequency Control Program. The automatic test feature shall be verified OPERABLE at the frequency specified in the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or other specified conditions for surveillance testing of the following:

a. Pressurizer Pressure Safety Injection Automatic Actuation Logic; and

b. Pressurizer Pressure Containment Isolation Automatic Actuation Logic; and

c. Steam Generator Pressure Main Steam Line Isolation Automatic Actuation Logic; and

d. Pressurizer Pressure Enclosure Building Filtration Automatic Actuation Logic.

Testing of the automatic actuation logic for Pressurizer Pressure Safety Injection, Pressurizer Pressure Containment Isolation, and Pressurizer Pressure Enclosure Building Filtration shall be performed within 12 hours after exceeding a pressurizer pressure of 1850 psia in MODE 3. Testing of the automatic actuation logic for Steam Generator Pressure Main Steam Line Isolation shall be performed within 12 hours after exceeding a steam generator pressure of 700 psia in MODE 3. MILLSTONE - UNIT 2 3/4 3-22 Amendment No. ii-,~. W, 324

INSTRUMENTATION ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SENSOR CABINET POWER SUPPLY DRAWERS LIMITING CONDITION FOR OPERATION 3.3.2.2 The engineered safety feature actuation system Sensor Cabinets (RC02Al, RC02B2, RC02C3 & RC02D4) Power Supply Drawers shall be OPERABLE and energized from the normal power source with the backup power source available. The normal and backup power sources for each sensor cabinet is detailed in Table 3.3-5a: CABINET NORMAL POWER BACKUP POWER RC02Al VA-10 VA-40 RC02B2 VA-20 VA-30 RC02C3 VA-30 VA-20 RC02D4 VA-40 VA-10 Table 3.3-5a APPLICABILITY: MODES I, 2, 3 and 4 ACTION: With any of the Sensor Cabinet Power Supply Drawers inoperable, or either the normal or backup power source not available as delineated in Table 3.3-5a, restore the inoperable Sensor Cabinet Power Supply Drawer to OPERABLE status within 48 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.3.2.2. l The engineered safety feature actuation system Sensor Cabinet Power Supply Drawers shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by visual inspection of the power supply drawer indicating lamps. 4.3.2.2.2 Verify the OPERABILITY of the Sensor Cabinet Power Supply auctioneering circuit at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 3-23 Amendment No . .J:-19., m, m, 324

INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3. l The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits. APPLICABILITY: As shown in Table 3.3-6. ACTION:

a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 2 hours or declare the channel inoperable.

b. With the number of OPERABLE channels less than the number of MINIMUM CHANNELS OPERABLE in Table 3.3-6, take the ACTION shown in Table 3.3-6.

The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.1.1 Each required radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-3. 4.3.3.1.2 DELETED 4.3 .3.1.3 Verify the response time of the control room isolation channel at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 3-24 Amendment No.~.~.~.'* m-,m,324

    ~                                                                   TABLE4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS b

r.n

    '"""3
    ~

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

      '   INSTRUMENT                                              CHECK        CALIBRATION            TEST                 REQUIRED
    ~     1.      AREA MONITORS N

a. Deleted

b. Control Room Isolation SFCP SFCP SFCP ALL MODES

c. Containment High Range SFCP SFCP* SFCP I, 2, 3, & 4 w

   ~
   "f     2.      PROCESS MONITORS N
   -....)

a. Containment Atmosphere- SFCP SFCP SFCP I, 2, 3, & 4 Particulate

b. Deleted

c. Noble Gas Effluent SFCP SFCP SFCP 1, 2, 3, & 4 Monitor (high range)

    >                     (Unit 2 Stack)
    ~
    ~ *

i Calibration of the sensor with a radioactive source need only be performed on the lowest range. Higher ranges may be calibrated electronically.

    ~

z

    ?
   ~t

~~~~ ~~~~ w N

INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room. APPLICABILITY: MODES 1, 2 and 3. ACTION: With the number of OPERABLE remote shutdown monitoring instrumentation channels less than required by Table 3.3-9, either:

a. Restore the inoperable channel to OPERABLE status within 7 days, or

b. Be in HOT SHUTDOWN within the next 24 hours.

SURVEILLANCE REQUIREMENTS 4.3.3.5 Each required remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6. MILLSTONE - UNIT 2 3/4 3-28 Amendment No. m, 324

~ TABLE4.3-6 (Z) ......, REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ~ tr:I CHANNEL CHANNEL I INSTRUMENT CHECK CALIBRATION

~        1. Wide Range Logarithmic Neutron Flux                SFCP      SFCP*
~

N

2. Reactor Trip Breaker Indication SFCP N.A.

3. Reactor Cold Leg Temperature SFCP SFCP

4. Pressurizer Pressure w a. Low Range SFCP SFCP

~ wI b. High Range SFCP SFCP w 0

5. Pressurizer Level SFCP SFCP

6. Steam Generator Level SFCP SFCP

7. Steam Generator Pressure SFCP SFCP

• Neutron detectors are excluded from the CHANNEL CALIBRATION.

>8 (I) g. 8 az 0 y~ ~w N

INSTRUMENTATION ACCIDENT MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:

a. ACTIONS per Table 3.3-11.

SURVEILLANCE REQUIREMENTS 4.3.3.8 Each required accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7. MILLSTONE - UNIT 2 314 3-31 Amendment No. 66, l:S+, m, m, 324

S::: TABLE 4.3-7

      ~
      ~                        ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
      ~                                                                         CHANNEL  CHANNEL
      ~ INSTRUMENT                                                               CHECK  CALIBRATION

J 1. Pressurizer Water Level SFCP SFCP N

2. Auxiliary Feedwater Flow Rate SFCP SFCP

3. Reactor Coolant System Subcooled/Superheat Monitor SFCP SFCP

(.;.) 4. PORV Position Indicator SFCP SFCP

     ~

(.;.) I (.;.)

5. PORV Block Valve Position Indicator N.A. SFCP Vi

6. Safety Valve Position Indicator SFCP SFCP

7. Containment Pressure SFCP SFCP

8. Containment Water Level (Narrow Range) SFCP SFCP

      ~
      =

9. Containment Water Level (Wide Range) SFCP SFCP sa (1)

10. Core Exit Thermocouples SFCP SFCP*

11. Main Steam Line Radiation Monitor SFCP SFCP z

      ~

12. Reactor Vessel Coolant Level SFCP SFCP*

u$:

     ~

u~; Electronic calibration from the ICC cabinets only.

     ~t

~~~ ~ N

REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.l.l Two reactor coolant loops shall be OPERABLE and in operation. APPLICABILITY: MODES 1 and 2. ACTION: With the requirements of the above specification not met, be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-1 Amendment No.~'@,~. 249, +/-9-1-, 324 Reissued by NRG Leiter de:tea September 27, 2006

REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 Two reactor coolant loops shall be OPERABLE and one reactor coolant loop shall be in operation. NOTE All reactor coolant pumps may not be in operation for up to 1 hour per 8 hour period provided:

a. no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1.1; and

b. core outlet temperature is maintained at least 10°F below saturation temperature.

APPLICABILITY: MODE 3. ACTION: a. With one reactor coolant loop inoperable, restore the required reactor coolant loop to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.

b. With no reactor coolant loop OPERABLE or in operation, immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1. l and immediately initiate corrective action to return one required reactor coolant loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 The required reactor coolant pump, if not in operation, shall be determined to be OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power available. 4.4.1.2.2 One reactor coolant loop shall be verified to be in operation at the frequency specified in the Surveillance Frequency Control Program. 4.4.1.2.3 Each steam generator secondary side water level shall be verified to be ;;:: 10% narrow range at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-la Amendment No. 69, ~. ~. 324

REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION HOT SHUTDOWN SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required pump, if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power available. 4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE, by verifying the secondary side water level to be 2: 10% narrow range at the frequency specified in the Surveillance Frequency Control Program. 4.4.1.3 .3 One reactor coolant loop or shutdown cooling train shall be verified to be in operation at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-lc Amendment No. 69, 249, 324

REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION COLD SHUTDOWN - REACTOR COOLANT SYSTEM LOOPS FILLED LIMITING CONDITION FOR OPERATION (continued) APPLICABILITY: MODE 5 with Reactor Coolant System loops filled. ACTION: a. With one shutdown cooling train inoperable and any steam generator secondary water level not within limits, immediately initiate action to either restore a second shutdown cooling train to OPERABLE status or restore steam generator secondary water levels to within limit.

b. With no shutdown cooling train OPERABLE or in operation, immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 and immediately initiate action to restore one shutdown cooling train to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1 The required shutdown cooling pump, if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power available. 4.4.1.4.2 The required steam generators shall be determined OPERABLE, by verifying the secondary side water level to be ~ 10% narrow range at the frequency specified in the Surveillance Frequency Control Program. 4.4. 1.4.3 One shutdown cooling train shall be verified to be in operation at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-le Amendment No. ~. m, 324

REACTOR COOLANT SYSTEM COOLANT LOOPS AND COOLANT CIRCULATION COLD SHUTDOWN REACTOR COOLANT SYSTEM LOOPS NOT FILLED M SURVEILLANCE REQUIREMENTS 4.4.1.5.1 The required shutdown cooling pump, if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power available. 4.4.1.5.2 One shutdown cooling train shall be verified to be in operation at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE UNIT 2 M Amendment No.~' 324

REACTOR COOLANT SYSTEM COLD SHUTDOWN - REACTOR COOLANT PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.6 A maximum of two reactor coolant pumps shall be OPERABLE. APPLICABILITY: MODE 5 ACTION: With more than two reactor coolant pumps OPERABLE, take immediate action to comply with Specification 3.4.1.6. SURVEILLANCE REQUIREMENTS 4.4.1.6 Two reactor coolant pumps shall be demonstrated inoperable at the frequency specified in the Surveillance Frequency Control Program by verifying that the motor circuit breakers have been disconnected from their electrical power supply circuits. MILLSTONE - UNIT 2 3/4 4-lh Amendment No. ~. ~, 324

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and

b. At the frequency specified in the Surveillance Frequency Control Program by performance of a CHANNEL CALIBRATION.

c. At the frequency specified in the Surveillance Frequency Control Program by operating the PORV through one complete cycle of full travel at conditions representative of MODES 3 or 4.

4.4.3.2 Each block valve shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by operating the valve through one complete cycle of full travel. This demonstration is not required if a PORV block valve is closed in accordance with the ACTIONS of Specification 3.4.3. MILLSTONE - UNIT 2 3/4 4-3a Amendment No. 66, 6%, ~. ~' 3-14, 324

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with:

a. Pressurizer water level ~ 70%, and

b. At least two groups of pressurizer heaters each having a capacity of at least 130kW.

APPLICABILITY: MODES 1, 2 and 3. ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 12 hours.

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours.

SURVEILLANCE REQUIREMENTS 4.4.4.1 The pressurizer water level shall be determined to be within its limits at the frequency specified in the Surveillance Frequency Control Program. 4.4.4.2 Verify at least two groups of pressurizer heaters each have a capacity of at least 130 kW at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-4 Amendment No. 06, .'.74, fA., BG, +/-1-9,

                                                                                            ~,;!%,324

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

2. Appropriate grab samples of the containment atmosphere are obtained and analyzed for particulate radioactivity within 6 hours and at least once per 6 hours thereafter, and

3. A Reactor Coolant System water inventory balance is performed within 6 hours and at least once per 6 hours thereafter.

Otherwise, be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

a. Containment atmosphere particulate monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and

b. Containment sump level monitoring system-performance of CHANNEL CALIBRATION TEST at the frequency specified in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 3/4 4-8a Amendment~' 324

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System Operational LEAKAGE shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. I GPM UNIDENTIFIED LEAKAGE,

c. 75 GPD primary to secondary LEAKAGE through any one steam generator, and

d. 10 GPM IDENTIFIED LEAKAGE.

APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:

a. With any RCS operational LEAKAGE not within limits for reasons other than PRESSURE BOUNDARY LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within limits within 4 hours.

b. With ACTION and associated completion time of ACTION a. not met, or PRESSURE BOUNDARY LEAKAGE exists, or primary to secondary LEAKAGE not within limits, be in HOT STANDBY within 6 hours and be in COLD SHUTDOWN within 36 hours.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 - - - - - - - - - - - - - - - NOTES- - - - - - - - - - - - - - - - *

1. Not required to be performed until 12 hours after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-9 Amendment No.~. '!rt-,@, 83-, .J:-9-1.,

                                                                         ~* .J:..3.8.,~.~.~.324

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.2.2 - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - -

• Not required to be performed until 12 hours after establishment of steady state operation.

Verify primary to secondary LEAKAGE is s:; 75 gallons per day through any one SG at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-10 Amendment No. ~. m, 324

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8.1 Verify the specific activity of the primary coolant ~ 1100 µCi/gram DOSE EQUIVALENT XE-133 at the frequency specified in the Surveillance Frequency Control Program.* 4.4.8.2 Verify the specific activity of the primary coolant~ 1.0 µCi/gram DOSE EQUIVALENT 1-131 at the frequency specified in the Surveillance Frequency Control Program,* and between 2 and 6 hours after a THERMAL POWER change of

          ~ 15% RATED THERMAL POWER within a one hour period.

• Surveillance only required to be performed for MODE 1 operation, consistent with the provisions of Specification 4.0.1.

MILLSTONE - UNIT 2 3/4 4-14 Amendment No. -H-5, 3W, 324

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1

a. The Reactor Coolant System temperature and pressure shall be determined to be within the limits at the frequency specified in the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

b. DELETED MILLSTONE - UNIT 2 314 4-18 Amendment No. ~. :t;R, 324

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENT 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at the frequency specified in the Surveillance Frequency Control Program thereafter when the PORV is required OPERABLE.

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at the frequency specified in the Surveillance Frequency Control Program.

c. Verifying the PORV block valve is open at the frequency specified in the Surveillance Frequency Control Program when the PORV is being used for overpressure protection.

d. Testing in accordance with the inservice test requirements of Specification 4.0.5.

4.4.9.3.2 Verify no more than the maximum allowed number of charging pumps are capable of injecting into the RCS at the frequency specified in the Surveillance Frequency Control Program. 4.4.9.3.3 Verify no more than the maximum allowed number of HPSI pumps are capable of injecting into the RCS at the frequency specified in the Surveillance Frequency Control Program. 4.4.9.3.4 Verify the required RCS vent is open at the frequency specified in the Surveillance Frequency Control Program when the vent pathway is provided by vent valve(s) that is(are) locked, sealed, or otherwise secured in the open position, otherwise, verify the vent pathway at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 4-2lb Amendment No. M, +4+, ~. U-8,

                                                                                        ~.~.324

EMERGENCY CORE COOLING SYSTEMS SAFETY INJECTION TANKS (Continued) SURVEILLANCE REQUIREMENTS 4.5.1 Each SIT shall be demonstrated OPERABLE:

a. Verify each SIT isolation valve is fully open at the frequency specified in the Surveillance Frequency Control Program.*(])

b. Verify borated water volume in each SIT is ~ 1080 cubic feet and~ 1190 cubic feet at the frequency specified in the Surveillance Frequency Control Program. **(2)

c. Verify nitrogen cover-pressure in each SIT is~ 200 psig and~ 250 psig at the frequency specified in the Surveillance Frequency Control Program.* n(3)

d. Verify boron concentration in each SIT is ~ 1720 ppm at the frequency specified in the Surveillance Frequency Control Program, and once within 6 hours after each solution volume increase of~ 1% of tank volume****(4) that is not the result of addition from the refueling water storage tank.

e. Verify that the closing coil in the valve breaker cubicle is removed at the frequency specified in the Surveillance Frequency Control Program.

• (1) If one SIT is inoperable, except as a result of boron concentration not within limits, or inoperable level or pressure instrumentation, surveillance is not applicable to the affected SIT.
  • (2) If one SIT is inoperable due solely to inoperable water level instrumentation, surveillance is not applicable to the affected SIT.
    • (3) If one SIT is inoperable due solely to inoperable pressure instrumentation, surveillance is not applicable to affected SIT.
      • (4)0nly required to be performed for affected SIT.

MILLSTONE - UNIT 2 314 5-2 Amendment No.#,~. m, US, 324

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying each Emergency Core Cooling System manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying that the following valves are in the indicated position with power to the valve operator removed:

Valve Number Valve Function Valve Position 2-SI-306 Shutdown Cooling Open* Flow Control 2-SI-659 SRAS Recirc. Open** 2-SI-660 SRAS Recirc. Open**

• Pinned and locked at preset throttle open position.

           **       To be closed prior to recirculation following LOCA.

c. By verifying the developed head of each high pressure safety injection pump at the flow test point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5.

d. By verifying the developed head of each low pressure safety injection pump at the flow test point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5.

e. By verifying the delivered flow of each charging pump at the required discharge pressure is greater than or equal to the required flow when tested pursuant to Specification 4.0.5.

f. At the frequency specified in the Surveillance Frequency Control Program by verifying each Emergency Core Cooling System automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

g. At the frequency specified in the Surveillance Frequency Control Program by verifying each high pressure safety injection pump and low pressure safety injection pump starts automatically on an actual or simulated actuation signal.

MILLSTONE - UNIT 2 3/4 5-4 Amendment No. ~. -W>, ~. m, 324

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

h. At the frequency specified in the Surveillance Frequency Control Program by verifying each low pressure safety injection pump stops automatically on an actual or simulated actuation signal.

i. By verifying the correct position of each electrical and/or mechanical position stop for each injection valve in Table 4.5-1:

1. Within 4 hours after completion of valve operations.

2. At the frequency specified in the Surveillance Frequency Control Program.

j. At the frequency specified in the Surveillance Frequency Control Program by verifying through visual inspection of the containment sump that each Emergency Core Cooling System subsystem suction inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion.

k. At the frequency specified in the Surveillance Frequency Control Program by verifying the Shutdown Cooling System open permissive interlock prevents the Shutdown Cooling System inlet isolation valves from being opened with an actual or simulated Reactor Coolant System pressure signal of~ 300 psia.

MILLSTONE - UNIT 2 314 5-5 Amendment No. +, #, ~, 6+, .f.G+,

                                                            +s9,+6.i.,~,~,~.~.~.324

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank shall be OPERABLE with:

a. A minimum contained volume of 370,000 gallons of borated water,

b. A minimum boron concentration of 1720 ppm,

c. A minimum water temperature of 50°F when in MODES 1 and 2, and

d. A minimum water temperature of35°F when in MODES 3 and 4.

APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the refueling water storage tank inoperable, restore tank to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying the water level in the tank, and

2. Verifying the boron concentration of the water.

b. When in MODES 3 and 4, at the frequency specified in the Surveillance Frequency Control Program by verifying the RWST temperature is;::: 35°F when the RWST ambient air temperature is < 35°F.

c. When in MODES 1 and 2, at the frequency specified in the Surveillance Frequency Control Program by verifying the RWST temperature is ;;:: 50°F when the RWST ambient air temperature is < 50°F.

MILLSTONE - UNIT 2 3/4 5-8 Amendment No. 324

EMERGENCY CORE COOLING SYSTEMS TRISODIUM PHOSPHATE (TSP) LIMITING CONDITION FOR OPERATION 3.5.5 The TSP baskets shall contain ~282 ft 3 of active TSP. APPLICABILITY: MODES 1, 2, and 3 ACTION: With the quantity of TSP less than required, restore the TSP quantity within 72 hours, or be in MODE 3 within the next 6 hours and MODE 4 within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.5.5.1 Verify that the TSP baskets contain ~282 ft3 of TSP at the frequency specified in the Surveillance Frequency Control Program. 4.5.5.2 Verify that a sample from the TSP baskets provides adequate pH adjustment of borated water at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 5-9 Amendment No.~, m, 324

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRJTY LIMITING CONDITION FOR OPERATION 3.6.l.1 Primary CONTAINMENT INTEGRJTY shall be maintained. APPLICABILITY: MODES I, 2, 3 and 4. ACTION: Without primary CONTAINMENT INTEGRJTY, restore CONTAINMENT INTEGRJTY within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that all penetrations(!) not caP,able of being closed by OPERABLE containment automatic isolation valves<2) and required to be closed during accident conditions are closed by: valves, blind flanges, or deactivated automatic valves secured in their positions,< 3) except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying the equipment hatch is closed and sealed.

c. By verifying the containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d. After each closing of a penetration subject to type B testing (except the containment air lock), if opened following a Type A or B test, by leak rate testing in accordance with the Containment Leakage Rate Testing Program.

e. By verifying Containment structural integrity in accordance with the Containment Tendon Surveillance Program.

( 1) Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise9 secured in the closed position. These penetrations shall be verified closed prior to entering MODE 4 from MODE 5, if not performed within the previous 92 days. (2) In MODE 4, the requirement for an OPERABLE containment automatic isolation valve system is satisfied by use of the containment isolation trip pushbuttons (3) Isolation devices in high radiation areas may be verified by use of administrative means. MILLSTONE - UNIT 2 3/4 6-1 Amendment No:~.%,~. m, ~.

                                                                                             *78,;!9.l-, 324

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS SURVEILLANCE REQUIREMENTS 4.6.1.3.1 Each containment air lock shall be demonstrated OPERABLE in accordance with the Containment Leakage Rate Testing Program. Containment air lock leakage test results shall be evaluated against the leakage limits of Technical Specification 3 .6.1.2. (An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test) 4.6.1.3 .2 Each containment air lock shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying that only one door in each air lock can be opened at a time. MILLSTONE - UNIT 2 3/4 6-6a Amendment No.-!*, ;!G;, 2:6+, 324

CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between -12 inches Water Gauge and +l.O PSIG. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containment internal pressure in excess of or below the limits above, restore the internal pressure to within the limits within 1 hour or be in HOT STANDBY within the next 4 hours; go to COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to within the limits at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 6-8 Amendment No.~' 324

CONTAINMENT SYSTEMS AIR 1EMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120°F. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containment average air temperature> 120°F, reduce the average air temperature to within the limit within 8 hours, or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be determined to be~ 120°F at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 6-9 Amendment No. U-9, 324

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY AND COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.2.1 Two containment spray trains and two containment cooling trains, with each cooling train consisting of two containment air recirculation and cooling units, shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3*. ACTION: Inoperable Equipment Required ACTION

a. Onecontamment a.I Restore the moperable containment spray train to spray train OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1750 psia within the following 6 hours.

b. One containment b.l Restore the inoperable containment cooling train to cooling train OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.

c. One containment c.l Restore the inoperable containment spray train or the spray train inoperable containment cooling train to OPERABLE status within 48 hours or be in HOT SHUTDOWN within the next AND 12 hours.

One containment cooling train

d. Two containment d.l Restore at least one inoperable containment cooling train to cooling trains OPERABLE status within 48 hours or be in HOT SHUTDOWN within the next 12 hours.

e. All other e.l Enter LCO 3.0.3 immediately.

combinations SURVEILLANCE REQUIREMENTS 4.6.2.1.1 Each containment spray train shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying each containment spray manual, power operated, and automatic valve in the spray train flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

• The Containment Spray System is not required to be OPERABLE in MODE 3 if pressurizer pressure is< 1750 psia.

MILLSTONE - UNIT 2 314 6-12 Amendment No.~.~' m, ~. mi-, 324

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. By verifying the developed head of each containment spray pump at the flow test point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5.

c. At the frequency specified in the Surveillance Frequency Control Program by verifying each automatic containment spray valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

d. At the frequency specified in the Surveillance Frequency Control Program by verifying each containment spray pump starts automatically on an actual or simulated actuation signal.

e. By verifying each spray nozzle is unobstructed following activities that could cause nozzle blockage.

4.6.2.1.2 Each containment air recirculation and cooling unit shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by operating each containment air recirculation and cooling unit in slow speed for~

15 minutes.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying each containment air recirculation and cooling unit cooling water flow rate is ~ 500 gpm.

c. At the frequency specified in the Surveillance Frequency Control Program by verifying each containment air recirculation and cooling unit starts automatically on an actual or simulated actuation signal.

MILLSTONE - UNIT 2 3/4 6-13 Amendment No.~. m, ~' 324

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3.1 Each containment isolation valve shall be OPERABLE.(l) (2) APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more of the isolation valve(s) inoperable, either:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours, or

b. Isolate the affected penetration(s) within 4 hours by use of a deactivated automatic valve(s) secured in the isolation position(s), or

c. Isolate the affected penetration(s) within 4 hours by use of a closed manual valve(s) or blind tlange(s); or

d. Isolate the affected penetration that has only one containment isolation valve and a closed system within 72 hours by use of at least one closed and deactivated automatic valve, closed manual valve, or blind flange; or

e. Be in COLD SHUTDOWN within the next 36 hours.

SURVEILLANCE REQUIREMENTS 4.6.3.1 Each containment isolation valve shall be demonstrated OPERABLE:

a. By verifying the isolation time of each power operated automatic containment isolation valve when tested pursuant to Specification 4.0.5.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

(1) Containment isolation valves may be opened on an intermittent basis under administrative controls. (2) The provisions of this Specification in MODES 1, 2 and 3, are not applicable for main steam line isolation valves. However, provisions of Specification 3. 7.1.5 are applicable for main steam line isolation valves. MILLSTONE - UNIT 2 3/4 6-15 Amendment No. 6, 2-1-G, m, m, 324

CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.3.2 The containment purge supply and exhaust isolation valves shall be sealed closed. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one containment purge supply and/or one exhaust isolation valve open, close the open valve(s) within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.3.2 The containment purge supply and exhaust isolation valves shall be determined sealed closed at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 6-19 Amendment No. 6+, ;;H6, 324

CONTAINMENT SYSTEMS POST-INCIDENT RECIRCULATION SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.4.4 Two separate and independent post-incident recirculation systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one post-incident recirculation system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.6.4.4 Each post-incident recirculation system shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by:

a. Verifying that the system can be started on operator action in the control room, and

b. Verifying that the system operates for at least 15 minutes.

MILLSTONE - UNIT 2 3/4 6-24 Amendment No. 324

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT ENCLOSURE BUILDING FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.1 Two separate and independent Enclosure Building Filtration Trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one Enclosure Building Filtration Train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.6.5.1 Each Enclosure Building Filtration Train shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 10 hours with the heaters on.

b. At the frequency specified in the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, and (2) following painting, fire or chemical release in any ventilation zone communicating with the train by:

MILLSTONE- UNIT 2 314 6-25 Amendment No. +/-G&, 324

CONTAINMENT SYSTEMS SURVEILLANCE"REQUIREMENTS (Continued)

1. Verifying that the cleanup 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 train flow rate is 9000 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, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

3. Verifying a train flow rate of 9000 cfin +/- 10% during train operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours 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, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978. *

d. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is ::;; 2.6 inches Water Gauge while operating the train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the train starts on an Enclosure Building Filtration Actuation Signal (EBFAS).

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% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 9000 cfm +/- 10%.

• ASTM 03803-89 shall be used in place of ANSI N509-l 976 as referenced in table 2 of Regulatory Guide 1.52. The laboratory test of charcoal should be conducted at a temperature of 30°C and a relative humidity of 95% within the tolerances specified by ASTM D3 803-89.

Additionally, the charcoal sample shall have a removal efficiency of~ 95%. MILLSTONE - UNIT 2 3/4 6-26 Amendment No.~. =R:, m, ~, m, 324

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

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% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N 510-197 5 while operating the train at a flow rate of 9000 cfm +/- 10%.

MILLSTONE - UNIT 2 314 6-27 Amendment No. 2-0&, 324

CONTAINMENT SYSTEMS ENCLOSURE BUILDING LI1\11TING CONDITION FOR OPERATION 3.6.5.2 The Enclosure Building shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the Enclosure Building inoperable, restore the Enclosure Building to OPERABLE status within 24 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.6.5.2.1 OPERABILITY of the Enclosure Building shall be demonstrated at the frequency specified in the Surveillance Frequency Control Program by verifying that each access opening is closed except when the access opening is being used for normal transit entry and exit. 4.6.5.2.2. At the frequency specified in the Surveillance Frequency Control Program verify each Enclosure Building Filtration Train produces a negative pressure of greater than or equal to 0.25 inches W.G. in the Enclosure Building Filtration Region within 1 minute after an Enclosure Building Filtration Actuation Signal. MILLSTONE - UNIT 2 3/4 6-28 Amendment No.~. 324

PLANT SYSTEMS AUXILIARY FEEDWATER PUMPS LIMITING CONDITION FOR OPERATION ACTION: (Continued) Inoperable Equipment Required ACTION

e. Three auxiliary feedwater e.

pumps in MODE 1, 2, or 3.

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

LCO 3.0.3 and all other LCO required ACTIONS requiring MODE changes are suspended until one AFW pump is restored to OPERABLE status. Immediately initiate ACTION to restore one auxiliary feedwater pump to OPERABLE status. SURVEILLANCE REQUIREMENTS

4. 7 .1.2 Each auxiliary feed water pump shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying each auxiliary feedwater manual, power operated, and automatic valve in each water flow path and in each steam supply flow path to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

b. By verifying the developed head of each auxiliary feedwater pump at the flow test point is greater than or equal to the required developed head when tested pursuant to Specification 4.0.5. (Not required to be performed for the steam turbine driven auxiliary feedwater pump until 24 hours after reaching 800 psig in the steam generators. The provisions of Specification 4.0.4 are not applicable to the steam turbine driven auxiliary feedwater pump for entry into MODE 3.)

MILLSTONE - UNIT 2 3/4 7-5 Amendment No.~.£,~. m, 324

PLANT SYSTEMS AUXILIARY FEEDWATER PUMPS SURVEILLANCE REQUIREMENTS (Continued)

c. At the frequency specified in the Surveillance Frequency Control Program by verifying each auxiliary feedwater automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position, as designed, on an actual or simulated actuation signal.

d. At the frequency specified in the Surveillance Frequency Control Program by verifying each auxiliary feedwater pump starts automatically, as designed, on an actual or simulated actuation signal.

e. By verifying the proper alignment of the required auxiliary feedwater flow paths by verifying flow from the condensate storage tank to each steam generator prior to entering MODE 2 whenever the unit has been in MODE 5, MODE 6, or defueled for a cumulative period of greater than 30 days.

MILLSTONE - UNIT 2 3/4 7-5a Amendment No. m, 324

PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank shall be OPERABLE with a minimum contained volume of 165,000 gallons. APPLICABILITY: MODES 1, 2 and 3. ACTION: With less than 165,000 gallons of water in the condensate storage tank, within 4 hours either:

a. Restore the water volume to within the limit or be in HOT SHUTDOWN within the next 12 hours, or

b. Demonstrate the OPERABILITY of the fire water system as a backup supply to the auxiliary feedwater pumps and restore the condensate storage tank water volume to within its limits within 7 days or be in HOT SHUTDOWN within the next 12 hours.

• SURVEILLANCE REQUIREMENTS 4.7.1.3 The condensate storage tank shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying the water level.

MILLSTONE - UNIT 2 3/4 7-6 Amendment No.~, 324

PLANT SYSTEMS ACTIVITY LIMITING CONDITION FOR OPERATION 3.7. I .4 The specific activity of the secondary coolant system shall be::; 0.10 uCi/gram DOSE EQUNALENT 1-131. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the specific activity of the secondary coolant system> 0.10 uCi/gram DOSE EQUIVALENT I-131, be in COLD SHUTDOWN within 36 hours after detection. SURVEILLANCE REQUIREMENTS

4. 7 .1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis of Table 4.7-2.

MILLSTONE - UNIT 2 3/4 7-7 Amendment No. 324

TABLE4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM AND ANALYSIS FREQUENCY

1. Gross Activity Determination At the frequency specified in the Surveillance Frequency Control Program.

2. Isotopic Analysis for DOSE a) 1 per 31 days, whenever the EQUIVALENT I-131 gross activity determination Concentration indicates iodine concentrations greater than 10% of the allowable limit b) At the frequency specified in the Surveillance Frequency Control Program, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit.

MILLSTONE - UNIT 2 3/4 7-8 Amendment No. ~.Mi:, 324

PLANT SYSTEMS MAIN FEEDWATER ISOLATION COMPONENTS (MFICs) LIMITING CONDITION FOR OPERATION (Continued)

b. With two or more of the feedwater isolation components inoperable in the same flow path, either:

1. Restore the inoperable component(s) to OPERABLE status within 8 hours until ACTION 'a' applies, or

2. Isolate the affected flow path within 8 hours, and verify that the inoperable feedwater isolation components are closed or isolated/secured once per 7 days, or

3. Be in HOT SHUTDOWN within the next 12 hours.

SURVEILLANCE REQUIREMENTS

4. 7 .1.6 Each feedwater isolation valve/feedwater pump trip circuitry shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by:

a. Verifying that on 'A' main steam isolation test signal, each isolation valve actuates to its isolation position, and

b. Verifying that on 'B' main steam isolation test signal, each isolation valve actuates to its isolation position, and

c. Verifying that on 'A' main steam isolation test signal, each feedwater pump trip circuit actuates, and

d. Verifying that on 'B' main steam isolation test signal; each feedwater pump trip circuit actuates.

MILLSTONE - UNIT 2 3/4 7-9b Amendment No. +88, m, 324 Reiss1:1ed by NRG Letter dated September 27, 2006

PLANT SYSTEMS ATMOSPHERIC DUMP VALVES LIMITING CONDITION FOR OPERATION 3.7.1.7 Each atmospheric dump valve line shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:

a. With one atmospheric dump valve line inoperable, restore the inoperable line to OPERABLE status within 48 hours or be in MODE 3 within the next 6 hours and MODE 4 within the following 24 hours.

b. With more than one atmospheric dump valve line inoperable, restore one inoperable line to OPERABLE status within 1 hour or be in MODE 3 within the next 6 hours and MODE 4 within the following 24 hours.

SURVEILLANCE REQUIREMENTS

4. 7. I. 7 Verify the OPERABILITY of each atmospheric dump valve line by local manual operation of each valve in the flowpath through one complete cycle of operation at the frequency specified in the Surveillance Frequency Control Program.

                                               - -- **- ----- - - -- -- -------- -- -- -- --------~---------

MILLSTONE - UNIT 2 3/4 7-9c Amendment No.~. m, 324

PLANT SYSTEMS STEAM GENERATOR BLOWDOWN ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.8 Each steam generator blowdown isolation valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3 ACTION: With one or more steam generator blowdown isolation valves inoperable, either:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours; or

b. Isolate the affected steam generator blowdown line within 4 hours; or

c. Be in MODE 3 within the next 6 hours and MODE 4 within the following 6 hours.

SURVEILLANCE REQUIREMENTS

4. 7.1.8 Verify the closure time of each steam generator blowdown isolation valve is ~ 10 seconds on an actual or simulated closure signal at the frequency specified in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 3/4 7-9d Amendment No.~' 324

PLANT SYSTEMS 3/4. 7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.l Two reactor building closed cooling water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one reactor building closed cooling water loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS

4. 7.3 .1 Each reactor building closed cooling water loop shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying each reactor building closed cooling water manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying each reactor building closed cooling water automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

c. At the frequency specified in the Surveillance Frequency Control Program by verifying each reactor building closed cooling water pump starts automatically on an actual or simulated actuation signal.

MILLSTONE - UNIT 2 3/4 7-11 Amendment No. m, ~. 324

PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 Two service water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one service water loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.7.4.l Each service water loop shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying each service water manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying each service water automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

c. At the frequency specified in the Surveillance Frequency Control Program by verifying each service water pump starts automatically on an actual or simulated actuation signal.

MILLSTONE - UNIT 2 3/4 7-12 Amendment No . .ffi, ~. m, 324

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.6.1 Each Control Room Emergency Ventilation Train shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that the control room air temperature is ::; 100°F.

b. At the frequency specified in the Surveillance Frequency Control Program by initiating from the control room, flow through the HEPA filters and charcoal adsorber train and verifying that the train operates for at least 15 minutes.

c. At the frequency specified in the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, and (2) following painting, fire or chemical release in any ventilation zone communicating with the train by:

1. Verifying that the cleanup 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 train flow rate is 2500 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, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

• The carbon sample shall have a removal efficiency of~ 95 percent.

3. Verifying a train flow rate of 2500 cfm +/- 10% during train operation when tested in accordance with ANSI N510-1975.

d. After every 720 hours 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, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978. *

• ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52. The laboratory test of charcoal should be conducted at a temperature of 30°C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

MILLSTONE - UNIT 2 3/4 7-17 Amendment No. +/-5, ti, +oo, H-9, -1+/-§.,

                                                                                       +49,+'.7.§.,~,324

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

e. At the frequency specified in the Surveillance Frequency Control Program by:

I. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorb er banks is less than 3 .4 inches Water Gauge while operating the train at a flow rate of 2500 cfin +/- 10%.

2. Verifying that on a recirculation signal, with the Control Room Emergency Ventilation Train operating in the normal mode and the smoke purge mode, the train automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks.

MILLSTONE - UNIT 2 3/47-17a Amendment No.~.~.~. H-9, ~.

                                                                                 -149, ~. ~' 324

PLANT SYSTEMS 3/4.7.11 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.11 The ultimate heat sink shall be OPERABLE with a water temperature ofless than or equal to 80°F. APPLICABILITY: MODES 1, 2, 3, AND 4 ACTION: With the UHS water temperature greater than 80°F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS

4. 7 .11 The ultimate heat sink shall be determined OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying the water temperature to be within limits.

b. At least once per 6 hours by verifying the water temperature to be within limits when the water temperature exceeds 75°F.

MILLSTONE - UNIT 2 3/4 7-34 Amendment No.~' -1:@, +9+, m,

                                                                                  ;wf,257,318,324

ELECTRICAL POWER SYSTEMS ACTION (Continued) Inoperable Equipment Required ACTION

e. Two diesel e.l Perform Surveillance Requirement 4.8.1.1.1 for the generators offsite circuits within 1 hour and at least once per 8 hours thereafter.

AND e.2 Restore one of the inoperable diesel generators to OPERABLE status within 2 hours or be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. AND e.3 Following restoration of one diesel generator restore remaining inoperable diesel generator to OPERABLE status following the time requirements of ACTION Statement b above based on the initial loss of the remaining inoperable diesel generator. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Verify correct breaker alignment and indicated power available for each required offsite circuit at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 314 8-2a Amendment No. +3-1-, ~. m., ~' 324

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.8. l .1.2 Each required diesel generator shall be demonstrated OPERABLE:*

a. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying the fuel level in the fuel oil supply tank, 2.

NOTES

1. A modified diesel generator start involving idling and gradual acceleration to synchronous speed may be used as recommended by the manufacturer. When modified start procedures are not used, the requirements of SR 4.8.1.1.2.d.l must be met.

2. Performance of SR 4.8.1.1.2.d satisfies this Surveillance Requirement.

Verifying the diesel generator starts from standby conditions and achieves steady state voltage~ 3740 V ands 4580 V, and Frequency~ 58.8 Hz and s 61.2 Hz. 3. NOTES

1. Diesel generator loading may include gradual loading as recommended by the manufacturer.

2. Momentary transients outside the load range do not invalidate this test.

3. This test shall be conducted on only one diesel generator at a time.

4. This test shall be preceded by and immediately follow without shutdown a successful performance of SR 4.8.1.1.2.a.2, or SRs 4.8.1.1.2.d.l and 4.8.1.1.2.d.2.

5. Performance of SR 4.8.1.1.2.d satisfies this Surveillance Requirement.

Verifying the diesel generator is synchronized and loaded, and operates for

                       ~ 60 minutes at a load~ 2475 kW ands 2750 kW.

• All diesel starts may be preceded by an engine prelube period.

MILLSTONE - UNIT 2 3/4 8-3 Amendment No. W, ~, m, 324

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. The diesel fuel oil supply shall be checked by:

1. Checking for and removing accumulated water from each fuel oil storage tank at the frequency specified in the Surveillance Frequency Control Program.

2. Verifying fuel oil properties of new and stored fuel oil are tested in accordance with, and maintained within the limits of, the Diesel Fuel Oil Testing Program in accordance with the Diesel Fuel Oil Testing Program.

c. At the frequency specified in the Surveillance Frequency Control Program by:

l. Deleted 2.

NOTE This surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, portions of the surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced. Verifying that the automatic time delay sequencer is OPERABLE with the following settings: Sequence Time After Closing of Diesel Generator Step Output Breaker (Seconds) Miriiriluiri ___ - 1 (T 1) 1.5 2.2 2 (T2) T 1 +5.5 8.4 3 (T3) T1+5.5 14.6 4 (T4) T3+5.5 20.8 MILLSTONE - UNIT 2 3/4 8-3a Amendment No. +3+, ~. ~. +/-++, 324

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENT (Continued)

d. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying the diesel starts from standby conditions and accelerates to

                  ~ 90% of rated speed and to ~ 97% of rated voltage within 15 seconds after the start signal.

2. Verifying the generator achieves steady state voltage~ 3740 V and

                  ~ 4580 V, and frequency ~ 58.8 Hz and ~ 61.2 Hz.

3. NOTES

1. Diesel generator loading may include gradual loading as recommended by the manufacturer.

2. Momentary transients outside the load range do not invalidate this test.

3. This test shall be conducted on only one diesel generator at a time.

4. This test shall be preceded by and immediately follow without shutdown a successful performance ofSRs 4.8.1.1.2.d.l and 4.8.1.1.2.d.2, or SR 4.8.1.1.2.a.2.

Verifying the diesel generator is synchronized and loaded, and operates for

                 ~  60 minutes at a load~ 2475 kW and~ 2750 kW.

MILLSTONE - UNIT 2 3/4 8-4 Amendment No. m, :;;;.:+, 324

ELECTRICAL POWER SYSTEMS 3/4 8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION

3. 8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators with tie breakers open between redundant busses:

4160 volt Emergency Bus # 24 C 4160 volt Emergency Bus #24 D 480 volt Emergency Load Center #22 E 480 volt Emergency Load Center #22 F 120 volt A.C. Vital Bus# VA-10 120 volt A.C. Vital Bus# VA-20 120 volt A.C. Vital Bus# VA-30 120 volt A.C. Vital Bus# VA-40 APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus and/ or associated load center to OPERABLE status within 8 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from normal A.C. sources with tie breakers open between redundant busses at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability. MILLSTONE - UNIT 2 314 8-6 AmendmentNo.~,324

ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION (Continued) 3.8.2. lA Inverters 5 and 6 shall be OPERABLE and available for automatic transfer via static switches VSl and VS2 to power busses VA-10 and VA-20, respectively. APPLICABILITY: MODES 1, 2 & 3 ACTION:

a. With inverter 5 or 6 inoperable, restore the inverter to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.

b. With inverter 5 or 6 unavailable for automatic transfer via static switch VSl or VS2 to power bus VA-10 or VA-20, respectively, restore the automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours.

c. With inverters 5 and 6 inoperable or unavailable for automatic transfer via static switches VSl and VS2 to power busses VA-10 and VA-20, respectively, restore the inverters to OPERABLE status or restore their automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours.

SURVEILLANCE REQUIREMENTS 4.8.2.lA a. Verify correct inverter voltage, frequency, and alignment for automatic transfer via static switches VSl and VS2 to power busses VA-10 and VA-20, respectively, at the frequency specified in the Surveillance Frequency Control Program.

b. Verify that busses VA-IO and VA-20 automatically transfer to their alternate power sources, inverters 5 and 6, respectively, at the frequency specified in the Surveillance Frequency Control Program during shutdown.

MILLSTONE - UNIT 2 3/4 8-6a Amendment No.+%&, m, 324

ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel generator but aligned to an OPERABLE diesel generator: 1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Load Center 2 - 120 volt AC. Vital Busses APPLICABILITY: MODES 5 and 6. ACTION: With less than the above complement of AC. busses OPERABLE and energized, suspend all operations involving CORE ALTERATIONS and positive reactivity additions that could result in loss ofrequired SDM or boron concentration, and movement ofrecently irradiated fuel assemblies. SURVEILLANCE REQUIREMENTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from normal AC. sources at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability. MILLSTONE - UNIT 2 3/4 8-7 AmendmentNo . .J.9.+,~,~,324

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION -OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.3 125-volt D.C. bus Train A and 125-volt D.C. bus Train B electrical power subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one 125-volt D.C. bus train inoperable, restore the inoperable 125-volt D.C. bus train to OPERABLE status within 2 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.8.2.3. l Each 125-volt D.C. bus train shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability. 4.8.2.3.2 Each 125-volt D.C. battery bank and charger of Train A and Train B shall be demonstrated OPERABLE:

a. By verifying at the frequency specified in the Surveillance Frequency Control Program that the battery cell parameters meet Table 4.8-1 Category A limits.

b. By verifying at the frequency specified in the Surveillance Frequency Control Program the battery cell parameters meet Table 4.8-1 Category B limits.

MILLSTONE - UNIT 2 3/4 8-8 Amendment No. +GS,.:J..W, ?:f9, 324

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. At the frequency specified in the Surveillance Frequency Control Program by verifying that:

1. The cells, cell plates and battery racks show no visual indication of physical damage or deterioration that could degrade battery performance,

2. The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material, and

3. The battery charger will supply at least 400 amperes at a minimum of 130 volts for at least 12 hours.

d. At the frequency specified in 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 emergency loads for 8 hours when the battery is subjected to a battery service test.

e. At the frequency specified in 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.

MILLSTONE - UNIT 2 314 8-9 Amendment No. -J:G8., +&G, m, 324

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 One 125 - volt D.C. bus train electrical power subsystem shall be OPERABLE: APPLICABILITY: MODES 5 and 6. ACTION: With no 125-volt D.C. bus trains OPERABLE, suspend all operations involving CORE ALTERATIONS and positive reactivity additions that could result in loss of required SDM or boron concentration, and movement ofrecently irradiated fuel assemblies. SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. bus train shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability. 4.8.2.4.2 The above required 125-volt D.C. bus train battery bank and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2. MILLSTONE - UNIT 2 314 8-10 Amendment No. +&G, -!9f, m, ~.

                                                                                          ~,324

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION SYSTEMS (TURBINE BATTERY)- OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.5 The Turbine Battery 125-volt D.C. electrical power subsystem shall be OPERABLE. APPLICABILITY: MODES l, 2 & 3 ACTION:

a. With the Turbine Battery 125-volt D.C. electrical power subsystem inoperable, restore the subsystem to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.

SURVEILLANCE REQUIREMENTS 4.8.2.5.l Verify 125-volt D.C. bus 201D is OPERABLE at the frequency specified in the Surveillance Frequency Control Program. 4.8.2.5.2 125-volt D.C. battery bank 201D shall be demonstrated OPERABLE:

a. By verifying at the frequency specified in the Surveillance Frequency Control Program that the battery cell parameters meet Table 4.8-2 Category A limits.

b. By verifying at the frequency specified in the Surveillance Frequency Control Program the battery cell parameters meet Table 4.8-2 Category B limits.

c. At the frequency specified in the Surveillance Frequency Control Program by verifying that:

1. The cells, cell plates, and battery racks show no visual indication of physical damage or deterioration that could degrade battery perfonnance, and

2. The cell-to-cell and terminal connections are clean, tight, free of corrosion, and coated with anti-corrosion material.

d. At the frequency specified in 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 loads for 1 hour when the battery is subjected to a battery service test.

e. At the frequency specified in 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.

MILLSTONE - UNIT 2 3/4 8-11 Amendment No. +8-8, ~. 324

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATIONS LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained sufficient to ensure that the more restrictive of following reactivity conditions is met:

a. Either a Keff of 0.95 or less, or

b. A boron concentration of greater than or equal to 1720 ppm.

APPLICABILITY: MODE 6. NOTE Only applicable to the refueling canal when connected to the Reactor Coolant System ACTION: With the requirements of the above specification not satisfied, within 15 minutes suspend all operations involving CORE ALTERATIONS and positive reactivity additions and initiate and continue boration at greater than or equal to 40 gpm of boric acid solution at or greater than the required refueling water storage tank concentration (ppm) until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 1720 ppm, whichever is the more restrictive. 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 all filled portions of the reactor coolant system and the refueling canal shall be determined by chemical analysis at the frequency specified in the Surveillance Frequency Control Program. 4.9.1.3 Deleted MILLSTONE - UNIT 2 3/4 9-1 AmendmentNo.~,~.~.~.324

REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 Two source range neutron flux monitors shall be OPERABLE, 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, immediately suspend all operations involving CORE ALTERATIONS and operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3. 9 .1.

b. With both of the above required monitors inoperable, immediately initiate action to restore one monitor to OPERABLE status. Additionally, determine that the boron concentration of the Reactor Coolant System satisfies the requirements of LCO 3.9.l within 4 hours and at least once per 12 hours thereafter.

SURVEILLANCE REQUIREMENTS. 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. Deleted

b. A CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency- Control Program:-- - - - -- - -- --- - - --- - -- ------ -- - ----- - ---

c. A CHANNEL CHECK and verification of audible counts at the frequency specified in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 3/4 9-2 Amendment No. %3-, ~' 324

REFUELING OPERATIONS CONTAINMENT PENETRATIONS SURVEILLANCE REQUIREMENTS 4.9.4.1 Verify each required containment penetration is in the required status at the frequency specified in the Surveillance Frequency Control Program. 4.9.4.2 Deleted MILLSTONE - UNIT 2 3/4 9-5 Amendment No.~. 324

REFUELING OPERATIONS SHUTDOWN COOLING AND COOLANT CIRCULATION - HIGHWATER LEVEL LIMITING CONDITION FOR OPERATION ACTION: With no shutdown cooling train OPERABLE or in operation, perform the following actions:

a. Immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3. 9. 1 and the loading of irradiated fuel assemblies in the core; and

b. Immediately initate action to restore one shutdown cooling train to OPERABLE status and operation; and

c. Within 4 hours place the containment penetrations in the following status:

1. Close the equipment door and secure with at least four bolts; and

2. Close at least one personnel airlock door; and

3. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be closed with a manual or automatic isolation valve, blind flange, or equivalent.

SURVEILLANCE REQUIREMENTS 4.9.8.1 One shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate greater than or equal to 1000 gpm at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 9-8a Amendment No. +l-, ~. ~. ~.

                                                                                              ~,324

REFUELING OPERATIONS SHUTDOWN COOLING AND COOLANT CIRCULATION - LOW WATER LEVEL LIMITING CONDITION FOR OPERATION (continued)

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be closed with a manual or automatic isolation valve, blind flange, or equivalent.

SURVEILLANCE REQUIREMENTS 4.9.8.2.1 One shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate greater than or equal to 1000 gpm at the frequency specified in the Surveillance Frequency Control Program. 4.9.8.2.2 The required shutdown cooling pump, if not in operation, shall be determined OPERABLE at the frequency specified in the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power available. MILLSTONE - UNIT 2 3/4 9-8c Amendment No.~' 324

REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION

3. 9 .11 As a minimum, 23. 0 feet of water shall be maintained over the top of the reactor vessel flange.

APPLICABILITY: During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts. During movement of irradiated fuel assemblies within containment. ACTION: With the water level less than that specified above, immediately suspend CORE ALTERATIONS and immediately suspend movement of irradiated fuel assemblies within containment. SURVEILLANCE REQUIREMENTS 4.9 .11 The water level shall be determined to be within its minimum depth at the frequency specified in the Surveillance Frequency Control Program. MILLSTONE - UNIT 2 3/4 9-11 Amendment No. W, 324

REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION

3. 9.12 As a minimum, 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 STORAGE POOL. ACTION: With the requirement of the specification not satisfied, suspend all movement of fuel and spent fuel pool platform crane operations with loads in the fuel storage areas. SURVEILLANCE REQUIREMENTS

4. 9.12 The water level in the storage pool shall be determined to be within its minimum depth at the frequency specified in the Surveillance Frequency Control Program when irradiated fuel assemblies are in the fuel storage pool.

MILLSTONE - UNIT 2 3/4 9-12 Amendment No. 324

REFUELING OPERATIONS SHIELDED CASK LIMITING CONDITION FOR OPERATION 3 .9 .16 All fuel within a distance L from the center of the spent fuel pool cask laydown area shall have decayed for at least 90 days. The distance L equals the major dimension of the shielded cask. APPLICABILITY: Whenever a shielded cask is on the refueling floor. ACTION: With the requirements of the above specification not satisfied, do not move a shielded cask to the refueling floor. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.16 The decay time of all fuel within a distance L from the center of the spent fuel pool cask laydown area shall be determined to be ~ 90 days within 24 hours prior to moving a shielded cask to the refueling floor and at the frequency specified in the Surveillance Frequency Control Program thereafter. MILLSTONE UNIT 2w 314 9wl9 Amendment No.~' +G9, ~. ~.

                                                                                              ,;i.e,324

REFUELING OPERATIONS SPENT FUEL POOL BORON CONCENTRATION LIMITING CONDITION FOR OPERATION

3. 9 .17 The boron concentration in the spent fuel pool shall be greater than or equal to 1720 parts per million (ppm).

APPLICAB1LITY: Whenever any fuel assembly or consolidated fuel storage box, is stored in the spent fuel pool. ACTION: With the boron concentration less than 1720 ppm, suspend the movement of all fuel, consolidated fuel storage boxes, and shielded casks, and immediately initiate action to restore the spent fuel pool boron concentration to within its limit. The provisions of specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.17 Verify that the boron concentration is greater than or equal to 1720 ppm at the frequency specified in the Surveillance Frequency Control Program, and within 24 hours prior to the initial movement of a fuel assembly or consolidated fuel storage box in the Spent Fuel Pool, or shielded cask over the cask laydown area. MILLSTONE - UNIT 2 314 9-21 Amendment No. +G9, W, 58, ~. m,324

SPECIAL TEST EXCEPTIONS GROUP HEIGHT AND INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The requirements of Specifications 3.1.1.4, 3.1.3.l, 3.1.3.5, 3.l.3.6, 3.2.3 and 3.2.4 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 below.

APPLICABILITY: MODES 1 and 2. ACTION: With any of the limits of Specification 3.2.l being exceeded while the requirements of Specifications 3.1.1.4, 3.1.3.l, 3.1.3.5, 3.l.3.6, 3.2.3 and 3.2.4 are suspended, immediately:

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1 or

b. Be in HOT STANDBY within 2 hours.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at the frequency specified in the Surveillance Frequency Control Program during PHYSICS TESTS in which the requirements of Specifications 3. l.l.4, 3.l.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 or 3.2.4 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 Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 or 3.2.4 are suspended. MILLSTONE - UNIT 2 3/4 10-2 Amendment No. ~* .£, +3-9, ~' 324

ADMINISTRATIVE CONTROLS 6.27 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (Continued)

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 inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses ofDBA consequences. Unfiltered air inleakage 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 Surveillance Requirement 4.0.2 are applicable to the frequencies for assessing CRE habitability and determining CRE unfiltered inleakage as required by paragraph c.

6.28 SNUBBER EXAMINATION. 1ESTING. AND SERVICE LIFE MONITORING PROGRAM This program conforms to the examination, testing, and service life monitoring for dynamic restraints (snubbers) in accordance with 10 CFR 50.55a inservice inspection (ISi) requirements for supports. The program shall be in accordance with the following:

a. This program shall meet 10 CFR 50.55a(g) ISI requirements for supports.

b. The program shall meet the requirements for ISi of supports set forth in subsequent editions of the Code of Record and addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code and the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) that are incorporated by reference in 10 CFR 50.55a(b), subject to its limitations and modifications, and subject to Commission approval.

c. The program shall, as allowed by 10 CFR 50.55a(b)(3)(v), meet Subsection ISTA, "General Requirements" and Subsection ISTD, "Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants" in lieu of Section XI of the ASME BPV Code ISI requirements for snubbers, or meet authorized alternatives pursuant to 10 CFR 50.55a(a)(3).

d. The 120-month program updates shall be made in accordance with 10 CFR 50.55a (including 10 CFR 50.55a(b )(3)(v)) subject to the limitations and modifications listed therein.

6.29 SURVEILLANCE FREQUENCY CONTROL PROGRAM This program provides controls for surveillance frequencies. The program shall ensure that surveillance requirements specified in the technical specification are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. MILLSTONE - UNIT 2 6-33 Amendment No.~. *9, 324

ADMINISTRATIVE CONTROLS 6.29 SURVEILLANCE FREQUENCY CONTROL PROGRAM (Continued)

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.

MILLSTONE - UNIT 2 6-34 Amendment No.3241

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 324 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336

1.0 INTRODUCTION

By letter dated October 22, 2014 (Agencywide Document Access and Management System (ADAMS) Accession No. ML14301A112), as supplemented by letters dated June 5, July 20, and August 27, 2015 (ADAMS Accession Nos. ML15163A021, ML15205A341, and ML15246A124, respectively), Dominion Nuclear Connecticut, Inc. (the licensee) proposed changes to the Technical Specifications (TSs) for Millstone Power Station, Unit No. 2 (MPS2). The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 28, 2015 (80 FR 23601 ). The requested change is the adoption of NRG-approved Technical Specifications Task Force (TSTF) traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control- [Risk-lnformed Technical Specification Task Force (RITSTF)] Initiative 5b" (Reference 1). When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TS. All surveillance frequencies can be relocated except:

• Frequencies that reference other approved programs for the specific interval (such as the In-Service 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 hours after thermal power reaching :::: 95% RTP [rated thermal power]); and Enclosure 2
• Frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g., "drywell to suppression chamber differential pressure decrease").

The requested change includes the addition of a new program to TS Section 6, Administrative Controls as Specification 6.29. The new program is called the SFCP and describes the requirements for the program to control changes to the relocated surveillance frequencies. The TS Bases for each of the affected surveillance requirements are revised to state that the frequency is set in accordance with the SFCP. Some surveillance requirements Bases do not contain a discussion of the frequency. In these cases, the TS Bases describing the current frequency were added to maintain consistency with the Bases for similar surveillances. These instances are noted in the markup along with the source of the text. The proposed changes to TS Section 6, Administrative Controls, to incorporate the SFCP include a specific reference to Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," Revision 1 (Reference 2, hereafter referred as NEI 04-10) as the basis for making any changes to the surveillance frequencies once they are relocated out of the TS. In a letter dated September 19, 2007 (Reference 3), the NRC staff approved NEI 04-10, as acceptable for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, and the safety evaluation providing the basis for NRC acceptance of NEI 04-10.

2.0 REGULATORY EVALUATION

In the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" published in the Federal Register (58 FR 39132, July 22, 1993), the NRC addressed the use of Probabilistic Safety Analysis (PSA, currently referred to as Probabilistic Risk Assessment or PRA) in the Standard Technical Specifications. In discussing the use of PSA in Nuclear Power Plant TSs, the Commission wrote 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 10 CFR 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 two years later the NRC provided additional detail concerning the use of PRA in the "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement" published in the Federal Register (60 FR 42622, August 16, 1995). The Commission, in discussing the deterministic and probabilistic approach to regulation, and the Commission's extension and enhancement of traditional regulation, wrote in part: 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. The Commission provided its new policy, stating: Although PRA methods and information have thus far been used successfully in nuclear regulatory activities, there have been concerns that PRA methods are not consistently applied throughout the agency, that sufficient agency PRA/statistics expertise is not available, and that the Commission is not deriving full benefit from the large agency and industry investment in the developed risk assessment methods. 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. Implementation of the policy statement will improve the regulatory process in three areas: Foremost, through safety decision making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees. 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. In Title 10 of the Code of Federal Regulations (10 CFR) 50.36, 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) Surveillance requirements; (4) Design features; and (5) Administrative controls. These categories will remain in the MPS2 TSs. As stated in 10 CFR 50.36(c)(3), "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 addition of the SFCP to the TSs provides the necessary administrative controls to require that surveillances 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. Changes to surveillance frequencies in the SFCP are made using the methodology contained in NEI 04-10, 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 50, Appendix B, Criterion XVI, "Corrective Action," require licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. One of these actions may be to consider increasing the frequency at which a surveillance test is performed. In addition, the SFCP implementation guidance in NEI 04-10, requires monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs.

Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML100910006) 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" (ADAMS Accession No. ML100910008) 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" (ADAMS Accession No. ML070240001 and ML090410014) (References 6 and 7) 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 decision making for light water-reactors. General guidance for evaluating the technical basis for proposed risk-informed changes is provided in Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance" (ADAMS Accession No. ML071700658), of NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition." Guidance on evaluating PRA technical adequacy is provided in the SRP, Chapter 19, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial fuel Load" (ADAMS Accession No. ML12193A107). More specific guidance related to risk-informed TS changes is provided in SRP, Chapter 16, Section 16.1, "Risk-Informed Decisionmaking: Technical Specifications" (ADAMS Accession No. ML070380228), which includes changes to CTs [completion times] as part of risk-informed decision making. Section 19.2 of the SRP references the same criteria as RG 1.174 (Reference

4) and RG 1.177 (Reference 5) 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 CDF [core damage frequency] 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, for MPS2 provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the administrative controls of the TSs. TSTF-425, Revision 3, also requires the application of NEI 04-10, for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425, Revision 3, included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200. In accordance with NEI 04-10, PRA methods are used, in combination with plant performance data and other considerations, to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is in accordance with guidance provided in RG 1.174 and RG 1.177 in support of changes to surveillance test intervals. 3.1 Review Methodology RG 1.177 identifies five key safety principles required for risk-informed changes to TSs. Each of these principles is addressed by NEI 04-10. 3.1.1 Key Principle 1: The Proposed Change Meets Current Regulations The regulatory requirement of 10 CFR 50.36(c)(3) states that TSs will include surveillances which 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." Licensees are required by TS to perform surveillance tests, calibration, or inspection on specific safety-related system equipment (e.g., reactivity control, power distribution, electrical, and instrumentation) to verify system operability. Surveillance frequencies, currently identified in TSs, are based primarily upon deterministic methods such as engineering judgment, operating experience, and manufacturer's recommendations. The licensee's use of NRG-approved methodologies identified in NEI 04-10, provides a way to establish risk-informed surveillance frequencies that complements the deterministic approach and supports the NRC's traditional defense-in-depth philosophy. The surveillances themselves would remain 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 surveillance requirements are retained in TSs, but the related surveillance frequencies are relocated to licensee-controlled documents, such as surveillances performed in accordance with the In-Service 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 50, Appendix 8, and the monitoring required by NEI 04-10, 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 surveillance requirements specified in the TSs are performed at intervals sufficient to assure the

above regulatory requirements are met. Thus, this proposed change meets the first key safety principle of RG 1.177 by complying with current regulations. 3.1.2 Key Principle 2: The Proposed Change Is Consistent With the Defense-in-Depth Philosophy The defense-in-depth philosophy, the second key safety principle of RG 1.177, 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 are preserved, and the potential for the introduction of new common cause failure 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.

TSTF-425, Revision 3, requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. NEI 04-10 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 and RG 1.177 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 common cause failures (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. Both the quantitative risk analysis and the qualitative considerations assure 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. 3.1.3 Key Principle 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 plant licensing basis (including the Updated Final Safety Analysis Report and bases to TSs), since 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. Thus, safety margins are maintained by the proposed methodology, and the third key safety principle of RG 1.177 is satisfied. 3.1.4 Key Principle 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 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. TSTF-425, Revision 3, requires application of NEI 04-10 in the SFCP. NEI 04-10 satisfies the intent of RG 1.177 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 technical specifications for control of surveillance frequencies. 3.1.4.1 Quality of the PRA The quality of the MPS2 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 higher change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA. RG 1.200 provides regulatory guidance for assessing the technical adequacy of a PRA. Revision 2, the latest revision (Reference 7), of this RG endorses (with comments and qualifications) the use of the 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," (Reference 8), NEI 00-02, "PRA Peer Review Process Guidelines," (Reference 9) and NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard" (Reference 10). Revision 1 of this RG had endorsed the internal events PRA standard ASME RA-Sb-2005, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (Reference 11 ). For the internal events PRA, there are no significant technical differences in the standard requirements, and therefore assessments using the previously endorsed internal events standard are acceptable. The licensee has performed an assessment of the PRA models used to support the SFCP using the guidance of RG 1.200 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 II of the standard is required by NEI 04-10 forthe internal events PRA, and any identified deficiencies to those requirements are assessed further to determine any impacts of proposed decreases to surveillance frequencies, including the use of sensitivity studies where appropriate. A formal Industry PRA peer review of the MPS2 internal events PRA model was performed in 2000. All findings and observations (F&Os) from this peer review have been addressed except for one significance level B F&O. In addition, the licensee performed self-assessments of the MPS2 internal events PRA in 2007 and 2011, using the American Society of Mechanical Engineers (ASME) PRA Standard, ASME RA-Sb-2005 and ASME/ANS [American Nuclear Society] RA-Sa-2009, respectively. The licensee also performed a focused scope peer review on the Human Reliability (HR), Large Early Release Frequency (LERF), and Internal Flooding (IF) Supporting Requirements (SRs) of the ASME/ANS Standard RA-Sa-2009. The licensee addressed the "gaps" between their internal events PRA model and the PRA standard from the self-assessment and the focused scope peer review, and provided them in Table 1 of the LAR. The staff's evaluation of the risk significant F&Os is summarized below. Gap #1 (2000) for Supporting Requirement AS-10/AS-18. The F&O cites the lack of modeling of makeup to the condenser when the steam dump valves fail. The licensee stated that they will perform a sensitivity study, until the F&O is resolved, by adding the steam dump valves as a required support system for the MFW [main feedwater] function. In its June 5, 2015 response to NRC staff RAI 1, regarding resolution of this F&O which remains unresolved after 15 years, the licensee stated that they will resolve the F&O at the next PRA model update. Considering that the steam dump valves are a low risk impact to the PRA and they do not impact the TSTF-425 program, the staff concludes that the sensitivity study for this F&O is acceptable for this application. Gap #2 (Self-Assessment) for Supporting Requirement IE-A8. The F&O notes that the interview with plant personnel for potential initiating events has been overlooked by the licensee. The licensee stated that this is a documentation issue and has no impact on the result due to the interviews with the systems personnel that concluded no new initiating events were required to be added to the PRA model. The licensee further stated that this SR will remain unmet until they perform interviews with the operation personnel, in accordance with their PRA procedure. In its June 5, 2015 response to RAI 2, the licensee stated "[a]n interview between their PRA staff and a former MPS2 operations shift manager was conducted to determine if the current MSPS2 PRA model overlooked potential initiating events." The licensee found that no further initiating events were required. The staff concludes that, the licensee can adequately implement the NEI 04-10 guidance because they have properly considered the pertinent personnel, consistent with the SR. Gap #3 (Self-Assessment) for Supporting Requirement AS-A?. The F&O is related to the licensee not considering the time of adverse Moderator Temperature Coefficient in Anticipated Transient Without Scram. The licensee also did not model loss of seal cooling, loss of alternating current, inadvertent opening of Power Operated Relief Valves, and safety relief valves in all event tree models; and the omission of operator action that fails to throttle Auxiliary Feedwater after power restoration following a Station Blackout (SBO). The licensee stated that all of the issues have been resolved except for power restoration following a SBO. Until the

F&O is resolved, the licensee stated that they will perform a sensitivity study to model the restarts of required accident mitigation components. The staff concludes that the licensee has dispositioned this F&O for the application because they will perform a sensitivity study consistent with the NEI 04-10 guidance. Gap #4 (Self-Assessment) for Supporting Requirement AS-A 10. The F&O is related to the licensee not explaining how the differences in system requirements for each initiating event impact operator actions or system responses. The licensee stated that this F&O is a documentation issue and will not have any impact to the program. The licensee further stated that they will address this as a part of F&O HR-G4-01 (Gap #9). In its June 5, 2015 response to RAI 3, the licensee clarified that Gap #4 and Gap #9 do not have a direct relationship. The licensee further stated "the impact on operator actions or system responses ... has been adequately assessed in the MSP2 PRA model, but this process has not been adequately explained in the model documentation." The staff concludes that the licensee has dispositioned this F&O for the application, because the technical requirement is resolved and the documentation issue will not impact the program. Gap #5 (Self-Assessment) for Supporting Requirement AS-C2. The F&O relates to the licensee not documenting the one-to-one correlation between each initiating event, the associated event tree, the system success criteria, and associated basis. The peer review team also found that the licensee did not discuss accident sequences pending resolution of issues associated with AS-A? (Gap #3) and the licensee did not clearly explain operator actions and associated dependencies on system success. The licensee agreed that the documentation regarding the F&O needed to be completed. Because the peer review team did not identify any modeling deficiencies as a result of the documentation issue, this F&O is expected to have no impact on the TSTF-425 program. Gap #6 (Self-Assessment) for Supporting Requirement SY-A4. The F&O relates to the lack of documentation to indicate that the interviews and walkdowns were performed. The licensee stated that system engineers were interviewed to partially address the SR and will conduct additional interviews with operations personnel and walkdowns to close this F&O. The staff concludes that the licensee has adequately dispositioned this F&O for this application since the interviews with the system engineers were completed and the licensee's PRA procedures now require interviews and walkdowns. Gap #7 (Self-Assessment) for Supporting Requirement SY-A21. The F&O stated that the supporting room heatup calculations are not well documented, and failure of electrical load shedding and excessive humidity conditions that could lead to loss of function are not addressed. The licensee stated that the room heatup calculations have been performed for the most risk significant rooms. The licensee also added the failure of load shedding to the electric power fault tree and the DC [direct current] switchgear room cooling is modeled for equipment requiring DC power after the initiating event occurs. The licensee will perform a sensitivity study by including the loss of DC switchgear room chillers following a Turbine Building High Energy Line Break. The staff concludes that, the licensee has dispositioned this F&O for the application because they will perform a sensitivity study in accordance with NEI 04-10. Gap #8 (Focused Scope Peer Review) for Supporting Requirement HR-G3. The F&O identified the dependency factor and sigma sections of the Human Reliability Analysis calculator

worksheets as not being properly filled out. The licensee has corrected the calculator worksheets in draft format and will perform a sensitivity study with the corrected Human Error Probabilities (HEPs) until the corrected files are entered into the worksheets. The staff concludes that the licensee has dispositioned this F&O for the application because they will perform a sensitivity study, with corrected dependency factors and sigma values for each of the affected HEPs, in accordance with NEI 04-10. Gap #9 (Focused Scope Peer Review) for SR HR-G4. The F&O noted that the HEP timing information for HRA event OAADV1 showed two different times associated with the event (30 minutes for General Transient and 11 minutes for Loss of Main Feedwater). The peer review team further stated that the licensee should use 11 minutes since it is limiting and the 30 minutes may be non-conservative. The licensee will keep this F&O open until the issue is addressed, and in its June 5, 2015 response to RAI 4, the licensee plans to perform a sensitivity study with a combination of corrected HEPs. The staff concludes that the licensee has dispositioned this F&O for the application because the sensitivity study on the HEPs will be performed in accordance with NEI 04-10. Gap #1 O (Focused Scope Peer Review) for SRs LE-C2 and LE-C7. The F&O described issues with the licensee's calculation of the Severe Accident Mitigation Guidelines (SAMGs) operator action HEPs for the Steam Generator Tube Rupture scenario and unaddressed operator actions in the containment isolation failure analysis. The licensee will perform a sensitivity study by realistically modeling the SAMG operator action to feed a dry steam generator and also remove credit for closing the containment spray and safety injection MOVs in the containment isolation analysis. The staff concludes that the licensee has dispositioned this F&O for the application because they will perform sensitivity analyses to address the issues with the SRs consistent with NEI 04-10. Gap #11 (Focused Scope Peer Review) for SR LE-F1. The F&O was created because the licensee did not provide a quantitative evaluation and identification of the dominant LERF contributors to LERF by plant damage states. The licensee stated the dominant LERF contributors to LERF need to be presented by plant damage states, which requires enhancements to CAFTA LERF model but will not have an impact to the LERF results. In its June 5, 2015 response to RAI 6, the licensee reviewed the LERF results and described the significant LERF contributors. The staff concludes that the licensee's identification and evaluation of the significant LERF contributors has considered significant contributors consistent with Table 2-2.8-9 and that LERF can be quantified for the application since plant damage states provides another way to represent LERF results. Gaps #12 and #13 (Focused Scope Peer Review) for SRs IFPP-A4, IFSN-A14, and IFSN-A16. These F&Os found that the licensee's assumption of 30 inches and two hours for non-water tight doors is potentially non-conservative, and does not reflect the as-operated plan configuration. The licensee plans to use the water height door failure criteria from EPRI Report 1019194 or criteria generated from door failure calculations to address the 30 inches issue. Until the F&O is resolved, the licensee will perform a sensitivity study by using water height door failure criteria from EPRI Report 1019194. The staff concludes that the licensee has adequately dispositioned this F&O for the application because they will perform a sensitivity analysis, on the door height, consistent with NEI 04-10. The two hour isolation criteria is addressed below.

Gaps #13 and #15 (Focused Scope Peer Review) for SRs IFSN-A14, and IFSN-A16. These F&Os found that the licensee's qualitatively assumed a 2-hour isolation criteria for plant mitigative action, which requires justification to meet the SRs Capability Category II requirement. The licensee has revised the PRA procedure to better align with the criteria in the Standard and they plan to review the two hour isolation criteria against the revised procedure. The licensee further stated that if the appropriate justification is not provided, the model would be revised. Until the F&O is resolved, the licensee will perform a sensitivity study using the criteria in the revised PRA procedure. The staff concludes that the licensee has adequately dispositioned this F&O for the application because they will perform a sensitivity analysis, on the revised procedure, consistent with NEI 04-10. Gap #14 (Focused Scope Peer Review) for Supporting Requirement IFSN-A8. The peer review team found that the licensee did not identify inter-area flood propagation through areas connected via backflow through drain lines involving failed check valves and hatchways, which is explicitly defined in the PRA Standard. The licensee stated that they would perform a study to those pathways if new propagation or flood pathways were identified. In its June 5, 2015 response to RAI 7, the licensee conducted an investigation on inter-area propagation and identified that the scenarios are "bounded by currently modeled internal flood scenarios." The staff concludes that the licensee has dispositioned this F&O for the application because it made proper consideration of this potentially risk significant scenario by identifying inter-area propagation as described in the SR, and can adequately apply the TSTF-425 program. Gap #16 (Focused Scope Peer Review) for Supporting Requirement IFEV-A5. The F&O notes that the licensee's PRA model did not reflect the most recent pipe break frequencies. The licensee will include the latest available pipe rupture frequencies, found in EPRI Report 3002000079, and perform a sensitivity study using the latest available industry data in the EPRI Report. The staff concludes that the licensee has dispositioned this F&O for the application because they will perform a sensitivity analysis, with the latest available pipe rupture frequencies, consistent with NEI 04-10. Gap #17 (2009) for Supporting Requirement IFEV-A6. The F&O related to the use of only generic pipe rupture frequencies in the licensee's PRA model. The licensee stated, however, that it collected and considered plant specific information but found no adverse trends that required Bayesian updating of the generic pipe frequencies. Thus, the licensee characterized this as a documentation issue because the evaluation was not documented at the time of the peer review. The licensee stated that it will perform an analysis to confirm the results at the next PRA model update. The staff concludes that the licensee's review of plant specific information and impending model update to confirm the results of the original analysis is acceptable, and does not have an impact on the PRA model. Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the staff concludes that the level of PRA quality, combined with the proposed 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.

3.1.4.2 Scope of the PRA The licensee is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10 to determine its potential impact on risk (CDF and LERF) from internal events, fires, seismic, other external events, and shutdown conditions. In cases where a PRA of sufficient scope or quantitative risk models were 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. MPS2 has a full-scope PRA model, whose full-power internal events and internal flood portions have received a peer review, self-assessments, and focused scope peer reviews as discussed previously. MPS2 does not have a PRA model for internal fire events, external events, and shutdown conditions. In accordance with NEI 04-10, the licensee will perform an initial qualitative screening analysis, and if the qualitative information is not sufficient, it will perform a bounding analysis. The bounding analysis will be performed in accordance with NEI 04-10, Rev. 1, Step 1Ob, and it will be based on risk insights and analysis documented in the MPS2 Individual Plant Examination of External Events (IPEEE) report with consideration of the IPEEE accident sequences, as well as relevant operating experience and additional risk insights obtained since the IPEEE study, in the context of the current plant configuration and operation. The NRC staff finds this approach to be consistent with NEI 04-10, Step 10b guidance in performing a bounding analysis. 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. 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 is performed. The methodology adjusts the failure probability of the impacted SSCs, including any impacted common cause failure 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, and by sensitivity studies identified in NEI 04-10. The licensee's approach for the evaluations of the impact of selected testing strategy (i.e., staggered testing or sequential testing) is consistent with the guidance of NUREG/CR-6141 and NUREG/CR-5497, as discussed in NEI 04-10.

Thus, through the application of NEI 04-10, the MPS2 PRA modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance frequency, and is consistent with regulatory position 2.3.3 of RG 1.177. 3.1.4.4 Assumptions for Time Related Failure Contributions The failure probabilities of SSCs modeled in PRAs may include a separate standby time-related contribution and a cyclic demand-related contribution. NEI 04-10 criteria adjust the time-related failure contribution of SSCs affected by the proposed change to surveillance frequency. This is consistent with RG 1.177 Section 2.3.3 which permits separation of the failure rate contributions into demand and standby for evaluation of surveillance requirements. 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 reduced surveillance frequency, including reduced downtime, lesser potential for restoration errors, reduction of potential for test caused transients, and reduced test-caused wear of equipment, are identified qualitatively, but not quantitatively assessed. Thus, through the application of NEI 04-10, the licensee has employed reasonable assumptions with regard to extensions of surveillance test intervals, and is consistent with regulatory position 2.3.4 of RG 1.177. 3.1.4.5 Sensitivity and Uncertainty Analyses NEI 04-10 requires sensitivity studies to 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 capability category II of the PRA standard. Where the sensitivity analyses identify 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. Required monitoring and feedback of SSC performance once the revised surveillance frequencies are implemented will also be performed. Thus, through the application of NEI 04-10, 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. 3.1.4.6 Acceptance Guidelines The licensee will 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 the guidance contained in NRC approved NEI 04-10 in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1E-6 per year for change to CDF, and below 1E-7 per year for change to LERF. These are consistent with the acceptance criteria of RG 1.174 for very small changes in risk. Where the RG 1.174 acceptance criteria are not met, the process either considers revised surveillance frequencies which are consistent with RG 1.174 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. Otherwise, bounding quantitative analyses are required which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174 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 1E-5 per year for change to CDF, and less than 1E-6 per year for change to LERF, and the total CDF and total LERF must be reasonably shown to be less than 1E-4 per year and 1E-5 per year, respectively. These are consistent with the acceptance criteria of RG 1.174, as referenced by RG 1.177 for changes to surveillance frequencies. The 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 (less than 5E-8 CDF and 5E-9 LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies. The quantitative acceptance guidance of RG 1.174 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 continued reliability of the SSC's. The licensee's application of NEI 04-10 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. Therefore, the proposed methodology satisfies the fourth key safety principle of RG 1.177 by assuring any increase in risk is small consistent with the intent of the Commission's Safety Goal Policy Statement. 3.1.5 Key Principle 5: The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's adoption of TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 requires performance monitoring of SSCs whose surveillance frequency has 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 degradation of SSC performance, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may apply as part of the maintenance rule requirements. The performance monitoring and feedback specified in NEI 04-10 is sufficient to reasonably assure acceptable SSC performance and is consistent

with regulatory position 3.2 of RG 1.177. Thus, the fifth key safety principle of RG 1.177 is satisfied. 3.2 Addition of Surveillance Frequency Control Program to Administrative Controls The licensee has included the SFCP and specific requirements into the Administrative Controls, TS Section 6.29, Surveillance Frequency Control Program, as follows: 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 the 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.

The proposed program is consistent with the model application of TSTF-425, and is therefore acceptable. 3.3 Minor Clarification Change to Surveillance Requirements The licensee proposed to insert the word "required" to surveillance requirements 4.1.3.3, 4.3.1.1.1, 4.3.2.1.1, 4.3.3.1.1, 4.3.3.5, and 4.3.3.8. The licensee responded to the staff's request for additional information concerning the proposed changes via letter dated July 20, 2015 (ADAMS Accession No. ML15205A341 ). According to the licensee, the addition of the word to the surveillance requirements is" ... for clarification purposes only." The licensee further explained that: "The limiting conditions for operation (LCOs) associated with each of these SRs specify a minimum number of channels required to be operable in applicable modes of operation. Under certain conditions or modes, the LCOs allow less than the total number of channels to be operable (e.g., 2 out of 4 channels). As currently written, these SRs may be misleading since they imply that all channels are required to be demonstrated operable." This clarification is outside the scope of TSTF-425, Revision 3; however, using the SRP guidance, the staff has determined that the licensee's reasoning for the change is correct and that the changes are an acceptable clarification change to the language of the surveillance requirements. Therefore, including the change to the surveillance requirements continue to meet the requirements of 10 CFR 50.36(c)(3). 3.4 Summary and Conclusions The 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, identified in the administrative controls of the TS. The SFCP and TS Section 6.29 references NEI 04-10, 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 TS to a licensee-controlled document, provided that those frequencies are changed in accordance with NEI 04-10, which is specified in the Administrative Controls of the TS. The licensee's proposed adoption of TSTF-425 and risk-informed methodology of NEI 04-10 as referenced in the Administrative Controls of the TS, satisfies the key principles of risk-informed decision making applied to changes to the TS as delineated in RG 1.177 and RG 1.174, in that:

• The proposed change meets current regulations;
• The proposed change is consistent with defense-in-depth philosophy;
• 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.

Paragraph 50.36(c) of 10 CFR discusses the categories that will be included in TSs. Paragraph 50.36(c)(3) of 10 CFR discusses the specific category of Surveillance Requirements 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 10 CFR 50.36(c)(3).

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Connecticut State official was notified on October 2, 2015, 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 the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in 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 published in the Federal Register (FR) on April 28, 2015 (80 FR 23601) that the amendment involves no significant hazards consideration, and there has been no public comment on such finding.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," March 18, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML090850642).

2. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 58, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession Number ML071360456).

3. Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 58, Risk-Informed Method for Control of Surveillance Frequencies," September 19, 2007 (ADAMS Accession Number ML072570267).

4. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011 (ADAMS Accession Number ML100910006).

5. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," Revision 1, May 2011 (ADAMS Accession Number ML100910008).

6. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007 (ADAMS Accession Number ML070240001 ).

7. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009 (ADAMS Accession Number ML090410014 ).

8. ASME/ANS PRA Standard ASME/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."

9. NEI 00-02, Revision 1 "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision 1, May 2006 (ADAMS Accession Number ML061510621 ).

10. NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard," Revision 0, August 2006.

11. ASME PRA Standard ASME RA-Sb-2005, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications."

Principal Contributors: J. Evans D. Oneal Date: October 29, 2015

October 29, 2015 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM, ADOPTION OF TSTF-425, REVISION 3 (TAC NO. MF5096)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 324 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated October 22, 2014, as supplemented by letters dated June 5, July 20, and August 27, 2015. The amendment modifies the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee controlled program with the adoption of Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - [Risk-Informed Technical Specification Task Force (RITSTF)] Initiative Sb." Additionally, the amendment adds a new program, the Surveillance Frequency Control Program, to TS Section 6, Administrative Controls. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely,

                                               /RAJ Richard V. Guzman, Sr. Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336

Enclosures:

1. Amendment No. 324 to DPR-65

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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