ML15093A002

Issuance of Amendment No. 320, Delete Technical Specification Index and Make Other Editorial and Administrative Changes

ML15093A002
Person / Time
Site:	Millstone
Issue date:	05/20/2015
From:	Richard Guzman Plant Licensing Branch 1
To:	Heacock D Dominion Nuclear Connecticut
Guzman R
References
TAC MF3870
Download: ML15093A002 (23)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.

Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 May 20, 2015

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: ADMINISTRATIVE CHANGES AND CORRECTIONS TO THE TECHNICAL SPECIFICATIONS (TAC NO. MF3870)

Dear Mr. Heacock:

The Commission has issued the enclosed Amendment No. 320 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2. This amendment is in response to your application dated March 28, 2014.

The amendment deletes the Technical Specification (TS} Index and makes other editorial, corrective and minor changes to the TSs.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket No. 50-336

Enclosures:

1. Amendment No. 320 to DPR-65

2. Safety Evaluation cc w/encls: Distribution via ListServ Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

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. 320 Renewed License No. DPR-65

1.

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

A.

The application for amendment by the applicant dated March 28, 2014, 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;

8.

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.

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. 320, are hereby incorporated in the renewed operating 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 30 days of issuance.

Attachment:

Changes to the License and Technical Specifications Date of Issuance: May 20, 2015 FOR THE NUCLEAR REGULATORY COMMISSION Michael I. Dudek, Acting Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 320 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 3

Insert 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 Index, Pages I-XVIII 2-7 2-7 2-8 2-8 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 6-2 6-2 6-18a 6-18a 6-19 6-19 6-20 6-20 6-24 6-24 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. 320, 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. 320

PAGES I THROUGH XVIII HAVE BEEN INTENTIONALLY DELETED MILLSTONE - UNIT 2 I

Amendment No. 9, 8, -W4, -H+, +48,

+/-99, 3W-, 320

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MILLSTONE - UNIT 2 2-7 Amendment No. ~. £, +64;320

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TABLE 3.3-11 (Continued)

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 -

With the number of channels OPERABLE less than the MINIMUM CHANNELS OPERABLE, determine the subcooling margin once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 3 -

With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information, and monitor discharge pipe temperature once per shift to determine valve position. This ACTION is not required if the PORV block valve is closed with power removed in accordance with Specification 3.U.b or.\\A.3.l'.

ACTION 4 -

a.

With the number of OPERABLE accident monitoring instrumentation channels less than the total number of channels shown in Table 3.3-1 l, restore the inoperable channel(s) to OPERABLE status within 7 days, or submit a special report to the Commission pursuant to Specification 6. 9.2 within the next 10 days outlining the cause of the malfunction, the plans for restoring the channel(s) to OPERABLE status, and any alternate methods in affect for estimating the applicable parameter during the interim.

b.

With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or submit a special report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction, the plans for restoring the channel(s) to OPERABLE status, and any alternate methods in affect for estimating the applicable parameter during the interim.

MILLSTONE - UNIT 2 314 3-33 Amendment No. ~' m, ;w+, 320

ACTION 5 -

With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table.3.3-! !,

restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 -

With any channel of radiation monitoring instrumentation inoperable, portable hand-held radiation detection equipment will be used to assess radiation releases from the atmospheric dump valves and steam generator safeties subsequent to a steam generator tube rupture.

ACTION 7 -

Restore the inoperable system to OPERABLE status within 7 days or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. (See the ACTION statement in Technical Specification 3.4.6.1.).

ACTION 8 -

With the number of OPERABLE Channels one less than the MINIMUM CHANNELS OPERABLE in Table 3.3-1 l, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down or:

1.

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and

2.

Restore the system to OPERABLE status at the next scheduled refueling; and

3.

Initiate an alternate method of monitoring the Reactor Vessel inventory.

MILLSTONE - UNIT 2 314 3-34 Amendment No. HG, +49, ~,320

ADMINISTRATIVE CONTROLS FACILITY STAFF (CONTINUED)

d.

A radiation protection technician shall be on site when fuel is in the reactor. (Table 6.2-1)

e.

ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f.

Deleted 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N 18.1-1971

• for comparable positions. Exceptions to this requirement are specified in the Quality Assurance Program.

6.3.2 The operations manager or at least one operations middle manager shall hold a senior reactor operator license for Millstone Unit No. 2.

• As of November 1, 2001, applicants for reactor operator and senior reactor operator qualification shall meet or exceed the education and experience guidelines of Regulatory Guide 1.8, Revision 3, May 2000.

MILLSTONE - UNIT 2 6-2 Amendment No. -86, ~. -143-, +-18,.

+w, +9+ ~, 270, 288, 308,320

ADMINISTRATIVE CONTROLS CORE OPERATING LIMUS_l\\]':POR_l 6.9.1.8 a.

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

3/4.1.1.1 SHUTDOWN MARGIN (SDM) 3/4.1.1.4 Moderator Temperature Coefficient 3/4.1.3.6 Regulating CEA Insertion Limits 3/4.2.1 Linear Heat Rate 3/4.2.3 TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR-FTr 3/4.2.6 DNB Margin

b.

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

1)

EMF-96-029(P)(A) Volumes 1 and 2, "Reactor Analysis System for PWRs Volume 1 - Methodology Description, Volume 2 -Benchmarking Results,"

Siemens Power Corporation.

2)

ANF-84-73 Appendix B (P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels.

3)

XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company.

4)

XN-75-32(P)(A) Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company.

5)

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model S-RELAP5 Based," Framatome ANP.

6)

EMF-2087(P)(A), "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation.

7)

XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company.

MILLSTONE - UNIT 2 6-18a Amendment No. :36, 9-3-, +94, +++, +-1-S-,

++9, HG, H+/-, +48, +£, 69, ~.

~.~,;t6G.~,;;s+.~,320

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (CONT.)

8)

XN-NF-621(P)(A), "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company.

9)

XN-NF-82-06(P)(A), and Supplements 2, 4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company.

10)

ANF-88-l 33(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation.

11)

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company.

12)

ANF-89-15l(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation.

13)

EMF-1961 (P)(A), "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation.

14)

EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP.

15)

EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.

16)

EMF-92-116(P)(A) Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation.

17)

BAW-10240(P)(A) Revision 0, "Incorporation of M5' Properties in Framatome ANP Approved Methods," May 2004.

c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

d.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

MILLSTONE - UNIT 2 6-19 Amendment No. +48, ~. ~. ~

~. ~. ~. ~. ;+-9; 320

ADMINISTRATIVE CONTROLS STEAM GENERA TOR TUBE INSPECTION REPORT 6.9.1.9 A report shall be submitted within 180 days after initial entry into MODE 4 following completion of an inspection performed in accordance with TS 6.26, Steam Generator (SG)

Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

Degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each degradation mechanism,

f.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator.

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a.

Deleted

b.

Deleted

c.

Deleted

d.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

e.

Deleted

f.

Deleted

g.

RCS Overpressure Mitigation, Specification 3.4.9.3.

MILLSTONE - UNIT 2 6-20 Amendment No. 9, ~. -W4, H+, +48,

~. +6;,+9+, ~. ~'~'~'

~.

~.+/--99,;H.,320

ADMINISTRATIVE CONTROLS 6.15 RADIOLOGICAL EFFLUENT MONITORING AND OFF SITE DOSE CALCULATION MANUAL (REMODCM)

a.

The REMODCM shall contain the methodology and parameters used in the calculation of off site doses resulting from radioactive gaseous and liquid effiuents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b.

The REMODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification (i.9.1.6a and specification 6.9. l.6h.

Licensee initiated changes to the REMODCM:

a.

Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1.

sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2.

a determination that the change(s) will maintain the level of radioactive effiuent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I of 10 CFR 50, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b.

Shall become effective after review and acceptance by FSRC and the approval of the designated officer; and

c.

Shall be submitted to the Commission in the form of a complete, legible copy of the entire REMO DCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the REMO DCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

MILLSTONE - UNIT 2 6-24 Amendment No. 4;, i-6-3-, +69, +:74,

~' ~' ~, ~,320

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 320 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65

1.0 INTRODUCTION

DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 By letter dated March 28, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14093A027), the Dominion Nuclear Connecticut, Inc. (the licensee),

submitted a request for changes to the Millstone Power Station, Unit No. 2 (MPS2), Technical Specifications (TSs). The proposed changes would delete the TS Index and make other editorial, corrective and minor changes to the TSs.

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Nuclear Regulatory Commission's regulatory requirements related to the content of the TSs are contained in Title 10, Code of Federal Regulations (10 CFR), Part 50, Section 50.36, "Technical Specifications." The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and control settings, (2) limiting conditions for operation, (3) surveillance requirements (SRs), (4) design features, (5) administrative controls, (6) decommissioning, (7) initial notification, and (8) written reports.

Licensees are required by 10 CFR 50.54(1) to designate individuals to be responsible for directing the licensed activities of licensed operators. These individuals shall be licensed as SROs [senior reactor operators] pursuant to 10 CFR Part 55.

3.0 TECHNICAL EVALUATION

3.1 Deletion of TS index The licensee proposed to delete the entire MPS2 TS index, all pages from I through XVIII to eliminate the need to include TS index pages as part of the license amendment request (LAR) submittals for future TS changes and to issue revised index pages with an amendment. The TS index identifies the contents of the TSs and where the specific TS sections appear in the TSs but the index does not contain any technical information required by 10 CFR 50.36. This proposed change is editorial and does not involve any physical changes to structures, systems, or components (SSCs) in the plant, or the way SSCs are operated or controlled. Therefore, the proposed deletion is acceptable.

3.2 Correction to Figure 2-2-2 Proposed change:

Replace figure number "2-2-2" on TS Page 2-7 with "2.2-2."

The figure number on TS Page 2-7 for Figure 2-2-2, "Local Power Density - High Trip Setpoint Part 2 (OR2 Versus Y1)," is incorrect. This is a typographical error.

This proposed editorial change is corrective in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled. Therefore, the proposed change is acceptable.

3.3 Corrections to Figure 2-2-2 and 2.2-3 Proposed change:

Replace "Y 1" (Y sub 1) in two locations in Figure 2-2-2 (to be relabeled as Figure 2.2-2 as noted in section 3.2, above) and in four locations in Figure 2.2-3, "Thermal Margin/Low Pressure Trip Setpoint Part 1 (Y1 Versus A1)," with "Yi" (Y sub I).

There are six locations in these 2 figures where AXIAL SHAPE INDEX for the trip signals in the reactor protection system, is incorrectly labeled as "Y/' (Y sub 1). As shown in the Definitions section of the MPS2 TSs, the symbol for AXIAL SHAPE INDEX for the trip signals in the reactor protection system, is "Yi" (Y sub I).

This proposed editorial change is corrective in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled. Therefore, the proposed change is acceptable.

3.4 Correction to Table 3.3-11, "Accident Monitoring Instrumentation" ACTION 3 Proposed change:

Replace "3.4.3.a and 3.4.3.b" in TS Table 3.3-11, ACTION 3, with "3.4.3.b and 3.4.3.c."

On February 15, 1994, Amendment 185 to the MPS2 TSs (ADAMS Accession No. ML012920503) was approved by the NRC. This amendment added two new ACTIONS to TS 3.4.3, Reactor Coolant System Relief Valves, to address the power-operated relief valve (PORV) and block valve reliability concerns identified in Generic Letter 90-06, "Resolution of Generic Issue 70, 'Power-Operated Relief Valve and Block Valve Reliability,' and Generic Issue 94, 'Additional Low-Temperature Over Pressure Protection for Light-Water Reactors,'

Pursuant to 10 CFR 50.54(f)." As part of this change, the reference in Surveillance Requirement (SR) 4.4.3.2 was changed from Specification "3.4.3 a orb" to Specification "3.4.3 b or c" to coincide with the changes to TS 3.4.3 for when a PORV block valve is closed and power is removed. At the time of the submittal for Amendment 185, the reference in TS Table 3.3-11, ACTION 3, should have also been revised to be consistent with the changes made to the ACTIONS in TS 3.4.3.

The TS Table 3.3-11, ACTION 3 currently states:

This ACTION is not required if the PORV block valve is closed with power removed in accordance with Specification 3.4.3.a or 3.4.3.b.

This statement is contradictory since Specification 3.4.3.a refers to when power is maintained to the PORV block valve, not removed from the PORV block valve. Specifications 3.4.3.b and 3.4.3.c are the correct references when a block valve(s) is closed and power is removed. The licensee did not include the correct references due to oversight during the preparation of the LAR for Amendment 185.

This proposed change is editorial in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled. Therefore, the proposed change is acceptable.

3.5 Correction to Table 3.3-11, "Accident Monitoring Instrumentation" ACTION 5 A typographical error exists in TS Table 3.3-11, ACTION 5. Currently ACTION 5 states, in part:

".... or begin at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

The word "begin" is a typographical error and would be corrected as follows:

".... or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

The change is consistent with the wording used in other sections of the TSs and indicates what mode the plant should "be in." This proposed change is corrective in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled, and is, therefore, acceptable.

3.6 Revision of TS 6.3.2, "Facility Staff Qualifications" TS 6.3.2 currently states:

If the operations manager does not hold a senior reactor operator license for Millstone Unit No. 2, then the operations manager shall have held a senior reactor operator license at a Pressurized Water Reactor and an individual serving in the capacity of the assistant operations manager shall hold a senior reactor operator license for Millstone Unit No. 2.

The proposed change to TS 6.3.2 is as follows:

The operations manager or at least one operations middle manager shall hold a senior reactor operator license for Millstone Unit No. 2.

The TS 6.3.1 currently requires that each member of the facility staff shall meet or exceed the minimum qualifications of American National Standards Institute (ANSI) N18.1-1971, "Selection and Training of Nuclear Power Plant Personnel," for comparable positions (note that as of November 1, 2001, applicants for reactor operator and senior reactor operator (SRO) qualification shall meet or exceed the education and experience guidelines of Regulatory Guide 1.8, Revision 3, "Qualification and Training of Personnel for Nuclear Power Plants," May 2000 (available at ADAMS Accession No. ML003706932). The ANSI N18.1-1971 specifies that the operations manager hold an SRO license for the unit. In 1987, this guidance was revised in ANSI/ANS 3.1-1987 such that if the operations manager does not hold an SRO license, then the operations middle manager shall hold an SRO license.

The existing language used in TS 6.3.2, Facility Staff Qualifications, reflects a limited exception to ANSI N18.1-1971. Specifically, the exception allows the assistant operations manager (i.e.,

the operations middle manager referenced in ANSI/ANS 3.1-1987), if one is designated, to hold an SRO license if the operations manager does not hold an SRO license for the unit. This exception was approved for MPS2 under Amendment No. 178 (ADAMS Accession No. ML012850390). As stated in the NRC safety evaluation report related to Amendment No. 178, the NRC staff concluded that:

... the specification that either the Operations Manager or Assistant Operations Manager hold a Millstone Unit 2 SRO license, is consistent with the requirements of 10 CFR 50.54(1) and ensures that a licensed off-shift senior operator is directing the licensed activities of the licensed operators. Requiring an ANSI/ANS 3.1-1987 qualified and licensed Assistant Operations Manager when the Operations Manager does not hold a valid Millstone Unit 2 SRO license is consistent with the requirements of ANSI N 18.1-1971 and ensures there is site-specific detailed relevant technical and systems knowledge in a senior operations management position.

The proposed change is consistent with the NRG-approved limited exception to ANSI N18.1-1971. The licensee proposed to revise TS 6.3.2 such that the operations manager or at least one operations middle manager (e.g., an assistant operations manager or the supervisor in charge of the operations shift crews) shall hold an SRO license for MPS2.

Modifying the title in TS 6.3.2 to allow at least one operations middle manager (i.e., any position between the operations manager and the shift managers) to hold the SRO license on the unit meets the intent of ANSI N 18.1-1971. This change will allow the operations manager and other operations middle managers to perform higher level duties such as management, planning, and coordinating of operations activities with more general senior operator knowledge.

The proposed change to TS 6.3.2 is administrative in nature, pertains to manager qualifications and does not substantially change staffing requirements because the NRC staff finds that a middle manager and assistant operations manager are equivalent positions. This proposed change continues to be consistent with the requirements of 10 CFR 50.54(1) to ensure that a licensed off-shift senior operator is directing the licensed activities of the licensed operators and with the standards set forth in ANSI N18.1-1971 to ensure there is relevant operational experience and knowledge in senior operations management position(s). The TSs would continue to require the requisite management qualifications for safe operations at MPS2, and is, therefore, acceptable.

3. 7 Correction to TS 6. 9.1. 8 a. "Core Operating Limits Report" The TS 6.9.1.8 a currently states:

3/4.2.3 Total Integrated Radial Peaking Factor - FTr The proposed change is as follows:

3/4.2.3 TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR - FTr In TS 6.9.1.8 a, the word "unrodded" is incorrectly excluded from the term "Total Integrated Radial Peaking Factor - F\\" Additionally, since this term has a TS definition, it should be capitalized. This proposed editorial change is corrective in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled. Therefore, this proposed change is acceptable.

3.8 Corrections to TS 6.9.1.8 b, "Core Operating Limits Report" References The TS 6.9.1.8 b Reference 5 currently states:

ENF-2328(P)(A), "PWR Small Break LOCA Evaluation Model S-RELAP5 Based,"

Framatome ANP.

The proposed change to Reference 5 is as follows:

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model S-RELAP5 Based,"

Framatome ANP.

The proposed change to Reference 5 would correct a typographical error. The term "EFN" should be "EMF," which is the correct reference title for the Framatome ANP technical report.

The TS 6.9.1.8 b Reference 7 currently states:

XN-NF-44(NP)(A), "A Generic Analysis of the Control rod Ejection Transient for Pressurized water reactors," Exxon Nuclear Company.

The proposed change to Reference 7 is as follows:

XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company.

The proposed change to Reference 7 would correct several typographical errors by inserting "78," which is missing from the document number and capitalizing the first letter in "rod," "water

and "reactors" to be consistent with the format in the title of the reference.

The TS 6.9.1.8 b Reference 14 currently states:

EMF-2130(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP.

The proposed change to Reference 14 would correct the report number as follows:

EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP.

Each of these proposed editorial changes described above are corrective in nature and do not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled.

Therefore, these proposed changes are acceptable.

3.9 Correction to TS 6.9.2. "Special Reports" The TS 6.9.2 currently states, in part:

a.

RCS Overpressure Mitigation, Specification 3.4.9.3 The proposed change to TS 6.9.2 is as follows:

g.

RCS Overpressure Mitigation, Specification 3.4.9.3 The last item under TS 6.9.2 on Page 6-20 incorrectly is listed as "a" and should be "g".

This proposed editorial change is corrective in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled. Therefore, this proposed change is acceptable.

3.10 Correction to TS 6.15. "Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM)"

In TS 6.15, paragraph b under, "Licensee initiated changes to the REMODCM," currently states:

Shall become effective after review and acceptance by SORG [Site Operations Review Committee] and the approval of the designated officer; and The proposed change to revise TS 6.15, paragraph b under, "Licensee initiated changes to the REMODCM," is as follows:

Shall become effective after review and acceptance by FSRC and the approval of the designated officer; and The term SORC is no longer used at MPS2 as those functions are now performed by the Facility Safety Review Committee (FSRC). Accordingly, the acronym SORC has been replaced with FSRC. This proposed change updates the TS to include the current name of the committee, is corrective in nature, and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled, and is, therefore, acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Connecticut State official was notified on April 22, 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 installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register (FR) on November 25, 2014 (79 FR 70212). The amendment also involves changes to administrative procedures or requirements and makes editorial, corrective or other minor revisions. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). 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.

Principal Contributor: K. West Date: May 20, 2015

Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.

Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 May 20, 2015

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: ADMINISTRATIVE CHANGES AND CORRECTIONS TO THE TECHNICAL SPECIFICATIONS (TAC NO. MF3870)

Dear Mr. Heacock:

The Commission has issued the enclosed Amendment No. 320 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2. This amendment is in response to your application dated March 28, 2014.

The amendment deletes the Technical Specification (TS) Index and makes several other editorial, corrective and minor changes to the TSs.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket No. 50-336

Enclosures:

1. Amendment No. 320 to DPR-65

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

PUBLIC RidsAcrsAcnw _MailCTR Resource RidsNrrDorlDpr Resource RidsNrrSTSBResource RidsNrrLAKGoldstein Resource KWest, NRR ADAMS Accession No: ML15093A002 Sincerely, IRA/

Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation LPLl-1 R/F RidsNrrDssStsb Resource RidsNrrDorllpl1-1 Resource RidsNrrPMMillstone Resource RidsRgn1 MailCenter Resource

• SE memo dated OFFICE NRR/DORL/LPLl-1 /PM NRR/DORL/LPLl-1 /PM NRR/DORL/LPLl-1 /LA NRR/DSS/STSB/BC NAME BVenkataraman RGuzman KGoldstein RElliott*

DATE 4/06/2015 4/09/2015 4/06/2015 3/10/2015 OFFICE OGG NRR/DORL/LPLl-1 /BC(A)

NRR/DORL/LPLl-1 /PM NAME MYoung MDudek RGuzman DATE 5/12/2015 5/20/2015 5/20/2015