1CAN031401, License Amendment Request - Revised Pages Extend the Type a Test Frequency to 15 Years

License Amendment Request - Revised Pages Extend the Type a Test Frequency to 15 Years

ML14070A399
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/11/2014
From:	Jeremy G. Browning Entergy Operations
To:	Document Control Desk
References
1CAN031401
Download: ML14070A399 (7)

Text

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Jeremy G. Browning Vice President - Operations Arkansas Nuclear One 1CAN031401 March 11, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request - Revised Pages Extend the Type A Test Frequency to 15 Years Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

REFERENCE:

Entergy letter dated December 20, 2013, License Amendment Request Technical Specification Change to Extend the Type A Test Frequency to 15 Years (1CAN121302) (ML13358A195)

Dear Sir or Madam:

Entergy Operations, Inc. (Entergy) submitted a request for an amendment to Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS) to allow for the extension of the current the ten-year frequency of the ANO-1 Type A or Integrated Leak Rate Test (ILRT) that is required by TS 5.5.16 to be extended to 15 years on a permanent basis (Reference above).

During the acceptance review, the NRC noted that the marked up and clean pages of TS did not include all the proposed changes that were identified in Attachment 1 of the reference document. Specifically, the date of the last performed ILRT was not revised. In addition, the Staff requested a clarification to the words except that currently in the TS. The words except that give the appearance that the extension of the Type A Test is an exception to the requirements of NEI 94-01, Revision 3-A. NEI 94-01, Revision 3-A provides the justification to extend the frequency for this test. Entergy is not requesting any exceptions to NEI 94-01, Revision 3-A for ANO-1. The words except that are deleted and the remaining sentence is divided into two sentences. provides the revised markup pages of existing TS to show the proposed change. provides revised (clean) TS pages.

These changes do not alter the no significant hazards consideration (NSHC) determination that was provided in the reference document.

1CAN031401 Page 2 of 3 In addition, it was noted that Section 4.4 of Attachment 1 of the reference document referenced NEI 94-01, Revision 2. This is in error. It should have referenced NEI 94-01, Revision 3. This error does not impact the technical discussions provided previously.

In accordance with 10 CFR 50.91(b)(1), a copy of this submittal is being provided to the designated Arkansas state official.

If you have any questions or require additional information, please contact Stephenie Pyle at 479-858-4704.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on March 11, 2014.

Sincerely, Original Signed by Jeremy G. Browning JGB/rwc Attachments:

1. Proposed Technical Specification Changes (mark-up)

2. Revised (clean) Technical Specification Pages

1CAN031401 Page 3 of 3 cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Michael Orenak MS O-8G9A One White Flint North 11555 Rockville Pike Rockville, MD 20852 U. S. Nuclear Regulatory Commission Attn: Mr. Alan Wang MS O-8B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Attachment 1 to 1CAN031401 Proposed Technical Specification Changes (mark-up)

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.16 Reactor Building Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the reactor building as required by 10 CFR 50.54(o) and 10 CFR 50, 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 1995NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, dated July 2012., except that t The next Type A test performed after the December 16, 2005April 16, 1992 Type A test shall be performed no later than December 16, 2020 April 15, 2007.

In addition, the reactor building purge supply and exhaust isolation valves shall be leakage rate tested once prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days.

The peak calculated reactor building internal pressure for the design basis loss of coolant accident, Pa, is 54 psig.

The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.

Leakage rate acceptance criteria are:

a. Reactor Building leakage rate acceptance criteria is £ 1.0 La. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and 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 tested at 10 psig.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Reactor Building Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Reactor Building Leakage Rate Testing Program.

ANO-1 5.0-18 Amendment No. 215,219,228,239,242,

Attachment 2 to 1CAN031401 Revised (clean) Technical Specification Pages

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.16 Reactor Building Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the reactor building as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, dated July 2012. The next Type A test performed after the December 16, 2005 Type A test shall be performed no later than December 16, 2020.

In addition, the reactor building purge supply and exhaust isolation valves shall be leakage rate tested once prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days.

The peak calculated reactor building internal pressure for the design basis loss of coolant accident, Pa, is 54 psig.

The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.

Leakage rate acceptance criteria are:

a. Reactor Building leakage rate acceptance criteria is £ 1.0 La. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and 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 tested at 10 psig.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Reactor Building Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Reactor Building Leakage Rate Testing Program.

ANO-1 5.0-18 Amendment No. 215,219,228,239,242,