ML16068A312

Issuance of Amendment No. 326 to Revise Technical Specifications for Containment Leak Rate Testing

ML16068A312
Person / Time
Site:	Millstone
Issue date:	03/31/2016
From:	Richard Guzman Plant Licensing Branch 1
To:	Heacock D Dominion Nuclear Connecticut
Guzman R
References
CAC MF5838
Download: ML16068A312 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 31, 2016 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 TO REVISE TECHNICAL SPECIFICATIONS FOR CONTAINMENT LEAK RATE TESTING (CAC NO. MF5838)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 326 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 2, 2015, as supplemented by letter dated August 31, 2015.

The amendment revises the MPS2 Technical Specifications (TSs) to (1) align the peak calculated primary containment internal pressure (Pa) for the design basis loss of coolant accident in TS 6.19 to be consistent with the 10 CFR 50 Appendix, J, Option B definition of Pa, and (2) revise the acceptable methods of surveillance for leakage rate testing of the containment air lock door seals.

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,

~-v?~~

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. 326 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. 326 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 March 2, 2015, as supplemented by letter dated August 31, 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. 326, 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 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Travis L. Tate, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-65 Date of lssuanceMarch 31 , 2O1 6

ATTACHMENT TO LICENSE AMENDMENT NO. 326 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 page of the Appendix A Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert 3/4 6-2 3/4 6-2 3/4 6-6 3/4 6-6 6-26 6-26

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. 326, 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. 326

August 25, 2015 CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of< La, 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa*

b. A combined leakage rate of< 0.60 La for all penetrations and valves subject to Type B and C tests when pressurized to Pa*

c. A combined leakage rate of< 0.014 La for all penetrations that are secondary containment bypass leakage paths when pressurized to Pa*

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 La, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types Band C tests exceeding 0.60 La, or (c) with the combined bypass leakage rate exceeding 0.014 La, restore the leakage rate(s) to within the limit(s) prior to increasing the Reactor Coolant System temperature above 200°F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated in accordance with the Containment Leakage Rate Testing Program.

MILLSTONE - UNIT 2 3/4 6-2 Amendment No. H4, 56, +8-7, ~,

234, +/-98 326

August 25, 2015 CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 The containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of~ 0.05 La at Pa*

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

NUlE Entry and exit through the containment air lock door is permitted to perform repairs on the affected air lock components.

a. With one containment air lock door inoperable:

1. Verify the OPERABLE air lock door is closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4. Entry into an OPERATIONAL MODE or other specified condition under the provisions of Specification 3.0.4 shall not be made if the inner air lock door is inoperable.

b. With only the containment air lock interlock mechanism inoperable, verify an OPERABLE air lock door is closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and lock an OPERABLE air lock door closed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Verify an OPERABLE air lock door is locked closed at least once per 31 days there after. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. (Entry into and exit from containment is permissible under the control of a dedicated individual).

c. With the containment air lock inoperable, except as specified in ACTION a. or ACTION b. above, immediately initiate action to evaluate overall containment leakage rate per Specification 3 .6.1.2 and verify an air lock door is closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Restore the air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MILLSTONE - UNIT 2 3/4 6-6 Amendment No. 95, 5+, ~, W. 326

August 25, 2015 ADMINISTRATIVE CONTROLS 6.19 CONTAINMENT LEAKAGE RATE JESTING PROGRAM A program shall be established to implement the leakage rate testing of the primary containment as required by 10CFR50.54(o) and 10CFR50, Appendix J, Option Bas modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, "Industry Performance-Based Option of 10 CFR Part 50, Appendix J": The first Type A test performed after the June 10, 1995 Type A test shall be performed no later than June 10, 2010.

The peak calculated primary Containment internal pressure for the design basis loss of coolant accident is Pa* Pa is 53 psig. Containment leakage rate testing will be performed at the containment design pressure of 54 psig or higher.

The maximum allowable primary containment leakage rate, La, at Pa' is 0.5% of primary containment air weight per day.

Leakage rate acceptance criteria are:

a. Primary containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are< 0.60 La for the combined Type Band Type C tests, and< 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1. Overall air lock leakage rate is ::::; 0.05 La when tested at ~ Pa*

2. For each door, leakage rate is::::; 0.01 La when pressurized to~ 25 psig.

The provisions of SR 4.0.2 do not apply for test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of SR 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

6.20 RADIOACTIVE EFFLUENT CONTROLS PROGRAM This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the REMODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the REMODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10CFR 20.1001-20.2402; MILLSTONE - UNIT 2 6-26 Amendment No. +/-ffi-, ~, ~, ~,

w+326

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 326 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 application dated March 2, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15069A226), as supplemented by letter dated August 31, 2015 (ADAMS Accession No. ML15246A117), Dominion Nuclear Connecticut, Inc. (DNC, the licensee), submitted to the U.S. Nuclear Regulatory Commission (NRC, the Commission) a license amendment request (LAR) for changes to the technical specifications (TSs) for the Millstone Power Station, Unit 2 (MPS2).

The proposed amendment would revise the MPS2 TSs to (1) align the peak calculated primary containment internal pressure (Pa) for the design-basis loss-of-coolant accident (DBLOCA) in TS 6.19 to be consistent with the 10 CFR 50 Appendix, J, Option B definition of Pa, and (2) revise the acceptable methods of surveillance for leakage rate testing of the containment air lock door seals. Specifically, the amendment revises TSs 3.6.1.2, 3.6.1.3, and 6.19 to attain consistency within the TSs in the definition of Pa. Additionally, the acceptance criteria for leakage rate testing of containment air lock door seals in TS 6.19 would be revised by removing the requirement to perform the test using the pressure decay method as other available leakage rate testing methods would also be acceptable.

Subsequent to the March 2, 2015, LAR, DNC identified a more appropriate set of TS changes to align the MPS2 TSs with the 10 CFR 50 Appendix J, Option B definition of Pa, and submitted its modifications to the LAR, in supplemental letter dated August 31, 2015. Specifically, DNC proposes to delete the containment design pressure value of 54 psig from TSs 3.6.1.2.a and 3.6.1.3.b and add the numerical value of Pa to TS 6.19 to align the MPS2 TSs with that contained in 10 CFR 50 Appendix J, Option B. The supplement did not change the proposed revision to the acceptance criteria for leakage rate testing of containment air lock door seals.

Enclosure 2

The supplemental letter dated August 31, 2015, provided additional information that clarified the application and expanded the scope of the application as originally noticed in the Federal Register(FR), 80 FR 43126 on July 21, 2015. A subsequent notice was published in the FR on February 22, 2016, along with the NRC staff's proposed no significant hazards consideration determination (81 FR 8752).

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations, Section 50.36 (10 CFR 50.36), TSs, "establishes the regulatory requirements related to the contents of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. In accordance with 10 CFR 50.36(c)(2), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee must shutdown the reactor or follow any remedial action permitted by the TSs until the condition can be met. The licensee proposed changes to revise two existing LCOs, 3.6.1.2.a and 3.6.1.3.b.

In accordance with 10 CFR 50.54(0), primary reactor containments for water cooled power reactors shall be subject to the requirements set forth in Appendix J to 10 CFR Part 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J to 10 CFR Part 50, includes two options: "Option A - Prescriptive Requirements," and "Option B -

Performance-Based Requirements," either of which can be chosen for meeting the requirements of the Appendix. The testing requirements in 10 CFR Part 50, Appendix J ensure that (a) leakage through containments or systems and components penetrating containments does not exceed allowable leakage rates specified in the TS; and (b) integrity of the containment structure is maintained during the service life of the containment. Option A allows containment leak rate testing at lower than Pa with appropriate extrapolation of leak rates to Pa, whereas Option B requires the testing to be performed at Pa or higher and that Pa be identified in the TS. MPS2 has voluntarily adopted Option B for meeting the requirements of 10 CFR Part 50, Appendix J.

Option B of 10 CFR Part 50, Appendix J specifies performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by the performance of ( 1) Type A tests to measure the containment system overall integrated leakage rate; (2) Type B pneumatic tests to detect and measure local leakage rates across pressure-retaining leakage-limiting boundaries such as penetrations; and (3) Type C pneumatic tests to measure containment isolation valve leakage rates. After preoperational testing, these tests are required to be conducted at periodic intervals based on the historical performance of the overall containment system (for Type A tests), and based on the safety significance and historical performance of each boundary and isolation valve (for Type B and C tests) to ensure integrity of the overall containment system as a barrier to fission product release.

Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Rate Testing Program,"

dated September 1995 (ADAMS Accession No. ML003740058) endorses, with certain exemptions, Nuclear Energy Institute (NEI) Report 94-01, Revision 0, "Industry Guideline for Implementing Performance Based Option of 10 CFR Part 50, Appendix J" (ADAMS Accession

No. ML11327A025). NEI Report 94-01, Revision 0, provides guidance for implementing the Appendix J performance-based requirements and incorporates, by reference, the provisions of ANSl/ANS-56-8-1994, "Containment System Leakage Testing Requirements" (ADAMS Accession No. ML11327A024). RG 1.163 states that all technical methods and techniques for performing Type A, B, and C tests in ANSl/ANS-56.8-1994 are acceptable to the NRC staff.

Section 6 of ANSl/ANS-56.8-1994 describes acceptable test methods for leakage rate testing of containment boundaries (Type B tests) and isolation valves (Type C tests). The acceptable methods include, but are not limited to, pressure decay and makeup flow rate. Containment air lock leakage rate testing is categorized as a Type B test.

3.0 TECHNICAL EVALUATION

3.1 Pa and Containment Design Pressure 3.1.1 Proposed TS Changes Current TS 6.19, "Containment Leakage Rate Testing Program," states in part:

The peak calculated primary Containment internal pressure for the design basis loss of coolant accident is Pa.

Current LCO 3.6.1.2, "Containment Leakage," states in part:

Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of <La, 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, 54 psig.

Current LCO 3.6.1.3, "Containment Air Locks," states in part:

The containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of~ 0.05 of La at Pa (54 psig).

In its letter dated August 31, 2015, the licensee modified its March 2, 2015, LAR and proposed the following changes to the TSs (changes in bold):

TS 6.19, "Containment Leakage Rate Testing Program":

The peak calculated primary Containment internal pressure for the design basis loss of coolant accident is Pa. Pa is 53 psig. Containment leakage rate testing will be performed at the containment design pressure of 54 psig or higher.

LCO 3.6.1.2, "Containment Leakage,":

Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of <La, 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa. ,54 psig LCO 3.6.1., "Containment Air Locks,":

The containment airlock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of :s; 0.05 of La at Pa (54 psig).

3.1.2 NRC Staff Evaluation The definition of Pa in current TS 6.19 is consistent with the definition of Pa in Option B of 10 CFR Part 50, Appendix J, which states, "Pa (p.s.i.g) means the calculated peak containment internal pressure related to the DBLOCA as specified in the Technical Specifications." The current LCOs 3.6.1.2 and 3.6.1.3 can be interpreted as Pa being equal to 54 psig. The licensee stated that peak calculated containment pressure for the MPS2 LOCA is 52.5 psig and that 54 psig actually represents containment design pressure. The apparent conflation of Pa with design pressure in TS LCOs 3.6.1.2 and 3.6.1.3 existed since the time MPS2 received an operating license in 1975.

The NRC staff finds the proposed changes provide clarity between Pa and containment design pressure, and maintains consistency with the definition of Pa in 10 CFR 50, Appendix J, Option B. Rounding the calculated peak containment pressure from 52.5 psig to 53 psig and the requirement to perform the Appendix J leak testing at or above the design pressure of 54 psig will maintain the conservatism in the test results. The proposed change does not involve any physical alterations to systems, structures, or components used in the mitigation of any postulated accidents. The NRC staff further finds that the proposed changes are in nomenclature only and will not impact the frequency of Appendix J tests, containment leakage acceptance criteria, methods of leak testing or frequency of surveillance, or invalidate past leak test results and conclusions.

3.2 Leakage Rate Testing of Containment Air Lock Door Seals 3.2.1 Proposed TS Change Current TS 6.19, "Containment Leakage Rate Testing Program" states in part:

b. Air lock testing acceptance criteria are:

1. Overall air lock leakage rate is :s; 0.05 La when tested at<:: Pa.

2. For each door, pressure decay is :5 0.1 psig when pressurized to <::

25 psig for at least 15 minutes.

The licensee is proposing the following change to the TS (changes in bold):

b. Air lock testing acceptance criteria are:

1. Overall air lock leakage rate is :5 0.05 La when tested at 2: Pa.

2. For each door, leakage rate is :S 0.01 La pressure decay is ::5 0.1 psig when pressurized to <:: 25 psig for at least 15 minutes.

3.2.2 NRC Staff Evaluation Containment air lock leakage rate testing is categorized as a Type B test, typically used for seals, gaskets, etc. The licensee stated that it is the only Type B local leakage rate test at MPS2 using the pressure decay test. All other Type B and Type C tests are performed using the makeup flow method. The licensee provided the following reasons for requesting the proposed change:

• Pressure decay method for leakage testing of containment air lock seals is difficult and time consuming.

• Pressure decay method is a go or no go test which does not provide quantifiable leak rate information.

• Due to the small test volume, pressure decay can occur too fast for accurate measurement thus requiring multiple attempts to complete the test.

• The makeup flow method, which maintains a constant test pressure, is a more efficient test methodology that provides quantifiable leak rate information during testing.

Containment air lock leakage rate testing is categorized as a Type B test. The proposed revision to TS 6.19.b.2 is consistent with NUREG-1432, revision 4.0, "Standard Technical Specifications (STSs), Combustion Engineering Plants," Volume 1, Specifications (ADAMS Accession No. ML12102A165).

As stated in TS 6.19.b.1, the air lock leakage rate limit is 0.05 La, which is not being changed by the proposed revision to the TSs. The proposed surveillance limit of 0.01 La for the air lock door seal integrity check in TS 6.19.b.2 is well below the overall air lock leakage limit of 0.05 La and is consistent with the formatting in STS 5.5.16, "Containment Leakage Rate Testing Program," in NUREG-1432. The licensee stated its intention to use makeup flow method for door seal integrity tests. The deletion of the words "pressure decay is :5 0.1 psig" from TS 6.19.b.2 will allow the licensee the flexibility to use other acceptable methods in ANSl/ANS-568-1994, including the makeup flow method. The removal of the words "for at least 15 minutes" is also acceptable because they apply to pressure decay test method and is consistent with the

formatting in STS 5.5.16 in NUREG-1432. Therefore, the NRC staff finds the proposed change to TS 6.19.b.2 acceptable.

3.3 NRC Staff Conclusion

The NRC staff finds that the proposed changes to clarify the difference between Pa and the containment design pressure are acceptable. The definition of Pa is consistent with that of 10 CFR 50, Appendix J, Option 8. The related TS changes to TS 3.6.1.2 and TS 3.6.1.3 are changes in nomenclature only and do not impact the ability of the licensee to comply with Option 8, performance-based requirements of 10 CFR 50, Appendix J for MPS2. The NRC staff also finds the proposed change to test the containment air lock leakage rate with other acceptable methods in ANSl/ANS-568-1994, including the makeup flow method is acceptable because it is consistent with Option 8, performance-based requirements of 10 CFR 50, Appendix J.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Connecticut State official was notified on March 8, 2016, 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 no significant change in the types or significant increase in the amounts 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 (81 FR 8752). 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.

Principal Contributor: Nageswara R. Karipineni Date: March 31, 2016

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 TO REVISE TECHNICAL SPECIFICATIONS FOR CONTAINMENT LEAK RATE TESTING (CAC NO. MF5838)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 326 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated March 2, 2015, as supplemented by letter dated August 31, 2015.

The amendment revises the MPS2 Technical Specifications (TSs) to (1) align the peak calculated primary containment internal pressure (Pa) for the design basis loss of coolant accident in TS 6.19 to be consistent with the 10 CFR 50 Appendix, J, Option B definition of Pa, and (2) revise the acceptable methods of surveillance for leakage rate testing of the containment air lock door seals.

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, IRA/

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. 326 to DPR-65

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

PUBLIC RidsRgn1 Mail Center RidsNrrPMMillstone Resource RidsNrrLAKGoldstein RidsNrrDorlDpr Resource RidsAcrs_MailCenter Resource RidsNrrDeEpnb Resource RidsNrrDorlLpl 1-1 Resource N.Karipineni, NRR Accession No.: ML16068A312 *See memo dated February 3, 2016 OFFICE DORL/LPL 1-1/PM DORL/LPL 1-1/LA DE/EPNB/BC* DSS/STSB/BC NAME Kl..:ioldstem RGuzman (SFiqueroa for) RDennig SAndersen for RElliott DATE 3/10/16 3/10/16 2/3/16 3/14/16 OFFICE OGC DORL/LPL 1-1/BC DORL/LPL 1-1/PM K<.;;uzman NAME MYoung TTate (DPickett for)

DATE 3/28/16 3/30/16 3/31/16 Ut-t'ICIAL RECUt (LJl,;Ut-'Y