ML15069A226

License Amendment Request to Revise TS 6.19, Containment Leakage Rate Testing Program

ML15069A226
Person / Time
Site:	Millstone
Issue date:	03/02/2015
From:	Mark D. Sartain Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
15-027
Download: ML15069A226 (13)

Text

S.

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 IiDominioln Web Address: www.dom.com March 2, 2015 U.S. Nuclear Regulatory Commission Serial No 15-027 Attention: Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No. 50-336 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 LICENSE AMENDMENT REQUEST TO REVISE TS 6.19. CONTAINMENT LEAKAGE RATE TESTING PROGRAM In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc.

(DNC) is submitting a license amendment request to revise Technical Specification (TS) 6.19, Containment Leakage Rate Testing Program, for Millstone Power Station Unit 2 (MPS2). Specifically, DNC proposes to: 1) revise the definition of Pa in TS 6.19 that was introduced into the TSs in License Amendment 203 to be consistent with the Pa value in TSs 3.6.1.2 and 3.6.1.3, and 2) revise the acceptance criteria for leakage rate testing of containment air lock door seals to substitute the use of the makeup flow method in lieu of the pressure decay method currently used at MPS2.

Attachment 1 to this letter describes the proposed change and provides justification for the proposed change. Attachment 2 provides the marked-up TS page.

The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The Facility Safety Review Committee has reviewed and concurred with the determinations herein.

Issuance of this amendment is requested by March 2, 2016, with the amendment to be implemented within 60 days.

In accordance with 10 CFR 50.91(b), a copy of this license amendment request is being provided to the State of Connecticut.

Should you have any questions in regard to this submittal, please contact Wanda Craft at (804) 273-4687.

Sincerely, Mark D. Sartain Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this *Jday of ___r_ k ,2015.

My Commission Expires: £31A* 4 - 31* 2O16 _ CRAIy P LY Commonwealth of Virginia]

Reg. # 7518653 My Commission Expires December 31, 20L

'4(L

Serial No: 15-027 Docket No. 50-336 Page 2 of 2 Attachments:

1. Evaluation of Proposed License Amendment

2. Marked-Up Technical Specification Page Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 M. C. Thadani Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 B1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.15-027 Docket No. 50-336 ATTACHMENT I EVALUATION OF PROPOSED LICENSE AMENDMENT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 1 of 8 EVALUATION OF PROPOSED LICENSE AMENDMENT

1.0 INTRODUCTION

In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc.

(DNC) is submitting a license amendment request to revise Technical Specification (TS) 6.19, Containment Leakage Rate Testing Program, for Millstone Power Station Unit 2 (MPS2). Specifically, DNC proposes to: 1) revise the definition of Pa in TS 6.19 that was introduced into the TSs in License Amendment 203 to be consistent with the Pa value in TSs 3.6.1.2 and 3.6.1.3, and 2) revise the acceptance criteria for leakage rate testing of containment air lock door seals to substitute the use of the makeup flow method in lieu of the pressure decay method currently used at MPS2.

2.0 PROPOSED CHANGE

S DNC proposes to revise TS 6.19, Containment Leakage Rate Testing Program, to:

2.1 Revise the definition of Pa in TS 6.19 that was introduced into the TSs in License Amendment 203 to be consistent with the Pa value in TSs 3.6.1.2 and 3.6.1.3.

The second paragraph of TS 6.19 would be revised as follows (Note: Deleted text is struck-through and added text is italicized and bold):

The peak calculated primary Containment internal pressure for the design basis loss of coolant accident and main steam line break accidentis less than the containment design pressure of 54 psig, however, the containmentdesign pressureis conservatively selected as is Pa.

2.2 Revise the acceptance criteria for leakage rate testing of containment air lock door seals.

Specifically, DNC proposes to revise the leakage rate acceptance criteria associated with the use of the pressure decay methodology with the acceptance criteria associated with the use of the makeup flow method.

TS 6.19.b would be revised as follows (Note: Deleted text is struck-through and added text is italicized and bold):

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, prffeure decay is , 0. 1 psig leakage rate is < 0.01 La when pressurized to > 25 psig for at least 1 5 minutes.

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 2 of 8 A markup of the proposed TS 6.19 changes are provided in Attachment 2.

3.0 BACKGROUND

MPS2 is a Combustion Engineering pressurized water reactor with an atmospheric-type primary containment structure. The primary containment completely encloses the reactor, reactor coolant system, and portions of the auxiliary and engineered safety features systems. The primary containment is penetrated by access, piping, and electrical penetrations.

Primary containment integrity ensures that the release of radioactive materials from the containment atmosphere is restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitations, limits the site boundary radiation doses to within the limits of 10 CFR 50.67 during accident conditions.

The overall leak-tight integrity of the primary containment is verified by a Type A integrated leakage rate test as required by 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." The integrity of the primary containment penetrations and isolation valves is verified through Type B and Type C local leakage rate tests. These tests are performed to verify the essentially leak-tight characteristics of the primary containment at the design basis accident pressure.

The limitations on containment leakage rates for MPS2 are specified in TS 3.6.1.2.

These limits ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to < 0.75 La (maximum allowable containment leakage rate) during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

Surveillance testing for measuring containment leakage rates at MPS2 is in accordance with TS 6.19, "Containment Leakage Rate Testing Program." Currently, TS 6.19 requires that leakage rate testing of containment air lock door seals be performed using the pressure decay method.

4.0 TECHNICAL ANALYSIS

4.1 Revise the definition of Pa in TS 6.19 that was introduced into the TSs in License Amendment 203 to be consistent with the Pa value in TSs 3.6.1.2 and 3.6.1.3.

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 3 of 8 MPS2 received an initial operating license in 1975. Since that time, TS 3.6.1.2 and TS 3.6.1.3 have defined Pa as 54 pounds per square inch gauge (psig).

Unlike the Combustion Engineering standard TS plants which define Pa as the peak calculated containment pressure for the Loss of Coolant Accident (LOCA),

MPS2 TS 3.6.1.2 and TS 3.6.1.3 define Pa as the containment design pressure.

This containment design pressure bounds the peak calculated containment pressure for the Loss of Coolant Accident (LOCA) (52.5 psig) and the Main Steam Line Break (MSLB) (53.8 psig). The maximum allowable primary containment leakage rate, La (0.5% of the primary containment air weight per day), is used in the MPS2 Final Safety Analysis Report, Chapter 14, for the radiological dose calculations of both the LOCA and the MSLB. Basing the maximum allowable primary containment leakage rate, La, on the containment design pressure is conservative, since the leak rate at the peak calculated containment pressure of the LOCA or MSLB would be slightly lower.

Currently, MPS2 TS 6.19 incorrectly limits the definition of Pa as the peak calculated primary containment internal pressure for the design basis LOCA.

This definition, which was derived from the Combustion Engineering standard TSs, was incorrectly added to the MPS2 TSs in License Amendment 203 (Reference 9.1). The proposed change shown in Section 2.1 will correct this error.

4.2 Revise the acceptance criteria for leakage rate testing of containment air lock door seals.

Methods for determining containment leakage rates are specified in 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J allows licensees to choose primary containment leakage testing under Option A, "Prescriptive Requirements," or Option B, "Performance-Based Requirements." In September 1996, the NRC approved License Amendment 203 to allow the use of Option B for Type A, B, and C testing at MPS2.

As delineated in 10 CFR 50, Appendix J, specific guidance concerning a performance-based leakage test program, acceptable leakage rate test methods, procedures, and analyses that may be used to implement Option B requirements and criteria are provided in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program" (Reference 9.2). RG 1.163 states that:

"...licensees intending to comply with Option B in the amendment to Appendix J should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01, rather than using the test intervals specified in ANSI/ANS-56.8-1994. All other technical methods and techniques for performing Types A, B, and C tests contained in ANSI/ANS-56.8-1994 are acceptable to the NRC staff."

ANSI/ANS-56.8-1994, "Containment System Leakage Testing Requirements" (Reference 9.3), Section 6, describes the acceptable test methods for leakage

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 4 of 8 rate testing of containment boundaries and isolation valves (Type B and Type C tests). These methods include, but are not limited to, pressure decay and makeup flow rate.

Surveillance testing of the containment air lock seals provides assurance that the overall air lock leakage rate will not become excessive through continuous air lock use. Containment air lock leakage rate testing is categorized as a Type B test (for seals, gaskets, etc.) and is the only Type B local leakage rate test at MPS2 that is performed using the pressure decay method. The other Type B and Type C tests are performed using the makeup flow method.

At MPS2, use of the pressure decay method for leakage rate testing of containment air lock door seals has proven to be both difficult and time consuming. The pressure decay method is a go/no go test which does not provide quantifiable leak rate information. Additionally, due to the small test volume involved, the pressure drop in the volume can occur too fast for accurate measurement and often requires multiple attempts to complete the test. In contrast, the makeup flow method, which maintains a constant test pressure, is a more efficient test methodology that provides quantifiable leak rate information during containment air lock testing.

As permitted by 10 CFR 50, Appendix J, Option B, use of the makeup flow method provides a suitable and alternate means for determining the containment air lock leakage at MPS2. Additionally, this method is consistent with Combustion Engineering Standard TS plants using Option B (i.e., TS 5.5.16.d.2.b) as specified in NUREG-1432, Revision 4.0 (Reference 9.4).

The proposed change to TS 6.19 will allow use of the makeup flow method to verify seal integrity and permit direct comparison to the air lock leakage rate limit of 0.05 La. The surveillance limit of 0.01 La using the makeup flow method is well below the overall air lock leakage rate limit of 0.05 La and is consistent with the surveillance acceptance criteria contained in NUREG-1432. This proposed change will continue to provide assurance of containment air lock integrity and limit the site boundary radiation doses to within the limits of 10 CFR 50.67 during accident conditions.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration The NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 5 of 8 accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety. DNC has evaluated whether or not a significant hazards consideration (SHC) is involved with the proposed amendment. A discussion of these standards as they relate to this amendment request is provided below.

Criterion 1 Will operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

This proposed license amendment would revise the definition of Pa that was introduced into TS 6.19 under License Amendment 203 to be consistent with the Pa value in TSs 3.6.1.2 and 3.6.1.3. The design basis accident remains unchanged for the postulated events described in the MPS2 Final Safety Analysis Report (FSAR). Since the initial conditions and assumptions included in the safety analyses are unchanged, the consequences of the postulated events remain unchanged. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed amendment also revises the method of surveillance for leakage rate testing of the containment air lock door seals. The makeup flow method will continue to provide assurance that the containment leakage rate is within the limits assumed in the radiological consequences analysis of the design basis accident, therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2 Will operation of the facility in accordance with this proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment would: 1) revise the definition of Pa in TS 6.19 to be consistent with the Pa value in TSs 3.6.1.2 and 3.6.1.3, and 2) revise the method of surveillance for leakage rate testing of the containment air lock door seals.

The proposed amendment does not change the way the plant is operated and does not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed amendment. Similarly, the proposed amendment would not physically change any plant systems, structures, or components involved in the mitigation of any postulated accidents. Thus, no new initiators or precursors of a new or different kind of accident are created.

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 6 of 8 Furthermore, the proposed amendment does not create the possibility of a new failure mode associated with any equipment or personnel failures. Therefore, the proposed amendment would not create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3 Will operation of the facility in accordance with this proposed amendment involve a significant reduction in the margin of safety?

Response: No.

The proposed amendment would: 1) revise the definition of Pa in TS 6.19 to be consistent with the Pa value in TSs 3.6.1.2 and 3.6.1.3, and 2) revise the method of surveillance for leakage rate testing of the containment air lock door seals.

The proposed amendment does not represent any physical change to plant systems, structures, or components, or to procedures established for plant operation. The proposed amendment does not affect the inputs or assumptions of any of the design basis analyses and current design limits will continue to be met. Since the proposed amendment does not affect the assumptions or consequences of any accident previously analyzed, there is no significant reduction in the margin of safety.

Conclusion Based on the above, DNC concludes that the proposed amendment does not represent a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

5.2 Applicable Regulatory Requirements/Criteria On February 20, 1971, the Atomic Energy Commission published in the Federal Register the General Design Criteria [GDC] for Nuclear Power Plants. Although MPS2 was designed and licensed to the GDC, as issued on July, 11, 1967, DNC has attempted to comply with the intent of the newer GDC to the extent possible, recognizing previous design commitments. The following GDC are applicable to the change proposed herein:

Criterion 53 - Provisions for Containment Testing and Inspection The reactor containment is designed to permit (1) appropriate periodic inspection of the important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak tightness of penetrations which have resilient seals and expansion bellows.

Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36(c)(3),

"Surveillance requirements," provides the regulatory requirements related to test, calibration, or inspection to assure that the necessary quality of systems and

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 7 of 8 components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Methods for determining containment leakage are specified in 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."

Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, provides guidelines for leakage rate testing of containments when using the performance-based option, Option B, in 10 CFR 50, Appendix J.

6.0 ENVIRONMENTAL CONSIDERATION

DNC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined by 10 CFR 20, or an inspection or surveillance requirement. DNC has evaluated the proposed amendment and has determined that it does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the proposed amendment.

7.0 CONCLUSION

Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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.

8.0 PRECEDENTS 8.1 The proposed change to TS 6.19, "Containment Leakage Rate Testing Program,"

is similar to the license amendment requests approved for Sequoyah Units 1 and 2 on October 2, 1986 (ADAMS Accession No. ML013250363) and Millstone Power Station Unit 3 on September 26, 1988 (ADAMS Accession No. ML011780329).

Serial No.15-027 Docket No. 50-336 Attachment 1, Page 8 of 8 8.2 The proposed change to TS 6.19 is also consistent with the requirements contained in NUREG-1432, Revision 4.0, for Combustion Engineering Standard TS plants using Option B (i.e., TS 5.5.16.d.2.b).

9.0 REFERENCES

9.1 MPS2 License Amendment 203, dated September 20, 1996 (ADAMS Accession No. ML012920217).

9.2 Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,"

September, 1995.

9.3 ANSI/ANS-56.8-1994, "Containment System Leakage Testing Requirements,"

August 4, 1994.

9.4 NUREG-1432, Revision 4.0, "Standard Technical Specifications, Combustion Engineering Plants," Volume 1, Specifications.

Serial No.15-027 Docket No. 50-336 ATTACHMENT 2 MARKED-UP TECHNICAL SPECIFICATION PAGE DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.15-027 Docket No. 50-336 ADMINISTRATIVE CONTROLS 6.19 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the primary containment as required by 10CFR50.54(o) and IOCFR50, Appendix J, Option B as 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 NEt 94-01, Rev. 0, "Industry Perfornance-Based Option of 10 CFR Part 50, Appendix J': The first Type A test perforned after the June 10, 1995 Type A test shall be perforned no later than June 10, 2010.

The peak calculated primary Containment internal pressure for the design basis loss of coolant accident' The maximum a able primary containment leakage rate, La, at P, is 0.5% of primary containment air weight- day. and main steam line break accident is less than the Leaagerat ac *a*

eptne critria. -containment design pressure of 54 psig, however, Leakage rate acceptanc criteria. eriterithe containment design pressure is conservatively

a. Primary containment overall lea selected as unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and < 0.75 La for Type A tests;

.... . *. . ,*~leakage ratei <001L

b. Air lock testing acceptance criteria are: lke t . , L

1. Overall air lock leakage rate i 0.05 La whentested at > Pa"

2. For each door, pre..tre deey,. :9 - 0. -plyg when pressurized to >25 psig fers-el-leat 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. 2043, 2-50, 26, 2-,-5, 2-94-