2CAN070803, Supplement to License Amendment Request Regarding Technical Specification Changes to Relocate RCS Chemistry Requirements to TRM

Supplement to License Amendment Request Regarding Technical Specification Changes to Relocate RCS Chemistry Requirements to TRM

ML081850032
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/01/2008
From:	Mitchell T Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN070803, TAC MD8313
Download: ML081850032 (5)

Text

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Timothy G. Mitchell Vice President, Operations Arkansas Nuclear One 2CAN070803 July 1, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Supplement to License Amendment Request Regarding Technical Specification Changes To Relocate RCS Chemistry Requirements to TRM Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

REFERENCES:

1. Entergy letter dated March 13, 2008, License Amendment Request:

Technical Specification Changes To Relocate RCS Chemistry Requirements to TRM (2CAN030802) (TAC NO: MD8313)

Dear Sir or Madam:

By letter (Reference 1), Entergy Operations, Inc. (Entergy) proposed a change to the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specifications (TSs) to relocate the TS requirements for Reactor Coolant System (RCS) chemistry to the Technical Requirements Manual (TRM). As part of the submittal, an administrative change was proposed to TS page 3/4 4-14 to support removal of the chemistry pages, by designating in the page footer that the next page is 3/4 4-18.

During the final NRC review process of the submittal, the NRC reviewer noted a typographical error on TS page 3/4 4-14, namely in Limiting Condition for Operation (LCO) 3.4.6.2.e. This portion of the LCO was intended to refer to TS Table 3.4.6-1 (in two locations). One of these references, however, is incorrectly stated as Table 3.4.6.1. In order to facilitate the best possible product when the new TS page is approved, Entergy is submitting a revised TS page that corrects this error.

2CAN070803 Page 2 of 2 In addition, during correction of the aforementioned typographical error, Entergy noted that the reference to the next TS page should have been placed in the footer of TS page 3/4 4-14b. This sub-page to TS 3.4.6.2, Reactor Coolant System Operational Leakage, was not previously identified due to a filing system error. The new mark-up pages are included in attachment to this letter. The changes are administrative in nature and have no impact on the original No Significance Hazards Consideration included in Reference 1.

This letter contains no new commitments.

If you have any questions or require additional information, please contact Dale James at 479-858-4619.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 1, 2008.

Sincerely, TGM/dbb

Attachment:

Proposed Technical Specification Change (mark-up) cc: Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS O-7 D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health & Human Services P.O. Box 1437 Slot H-30 Little Rock, AR 72203-1437

Attachment to To 2CAN070803 Proposed Technical Specification Change (mark-up)

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

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary to secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6-.1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE or any primary to secondary leakage not within limit, be in at least HOT STANDBY within 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 />.

b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and primary to secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or 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 />.

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two valves* in each high pressure line having a non-functional valve and 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 />.

• These valves may include check valves for which the leakage rate has been verified, manual valves or automatic valves. Manual and automatic valves shall be tagged as closed to preclude inadvertent valve opening.

ARKANSAS - UNIT 2 3/4 4-14 Amendment No. 184,266, Order dated 4/20/81

TABLE 3.4.6-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES (a)(b)(c)

(CHECK VALVES)

System Check Valve No.

High-Pressure Safety Injection Loop A, cold leg 2SI-15A 2SI-13A Loop B, cold leg 2SI-15B 2SI-13B Loop C, cold leg 2SI-15C 2SI-13C Loop D, cold leg 2SI-15D 2SI-13D Low-Pressure Safety Injection Loop A, cold leg 2SI-14A Loop B, cold leg 2SI-14B Loop C, cold leg 2SI-14C Loop D, cold leg 2SI-14D NOTES (a) Maximum Allowable Leakage (each valve):

1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

(b) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

(c) Minimum test differential shall not be less than 150 psid.

ARKANSAS - UNIT 2 3/4 4-14b Amendment No.

Next Page is 3/4 4-18 Order Date 4/20/81