RA-09-029, Supplement to License Amendment Request Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and Clarification of A..

Supplement to License Amendment Request Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and Clarification of A..

ML090890777
Person / Time
Site:	Oyster Creek
Issue date:	03/30/2009
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-09-029
Download: ML090890777 (4)

Text

200 Exelon Way PA 19348 10 CFR 50.90 RA-09-029 March 30, 2009 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Supplement to License Amendment Request RE: Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and Clarification of a Frequency Example Using the Consolidated Line Item Improvement Process

Reference:

Letter from Pamela B. Cowan to U. S. NRC, "Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and Clarification of a Frequency Example Using the Consolidated Line Item Improvement Process,"

dated June 9, 2008 In the reference letter, Exelon Generation Company, LLC (Exelon), LLC and AmerGen Energy Company LLC (AmerGen) requested NRC approval to implement TSTF-475, Revision 1, "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action," for Clinton Power Station, Unit 1, Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, Oyster Creek Nuclear Generating Station, Peach Bottom Atomic Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2.

Subsequently, through several teleconferences between Exelon and the NRC's Exelon Fleet Project Manager, it was agreed that the proposed Technical Specification wording for Oyster Creek Nuclear Generating Station (OCNGS) would require a clarification. Specifically, it was agreed to replace the word "monthly" with the phrase "once per 31 days" in proposed Technical Specification 4.2.0. provides the OCNGS Technical Specification Marked-Up Page reflecting this proposed change.

Exelon has concluded that the supplemental proposed change does not require a change to the original No Significant Hazards Consideration contained in the referenced license amendment request. Pursuant to 10 CFR 50.91 (b)(1), a copy of this supplement is being provided to the designated officials of the State of New Jersey, as well as the chief executives of the township and county in which the facility is located.

U. S. Nuclear Regulatory Commission March 30, 2009 Page 2 of 2 No new regulatory commitments are established by this submittal.

If any additional information is needed, please contact Frank Mascitelli at (610) 765-5512.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30th day of March 2009.

Respectfully,

~f?t ~_

Pamela B. Cowan Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

1) Supplement to OCNGS Technical Specification Proposed Technical Specification Marked-Up Page 4.2-1 cc: S. J. Collins, Administrator, USNRC Region I M. S. Ferdas, USNRC Senior Resident Inspector, Oyster Creek C. Gratton, USNRC Project Manager, Exelon Fleet G. E. Miller, USNRC Project Manager, Oyster Creek Mayor of Lacey Township P. Baldauf, Assistant Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection

Attachment 1 Supplement to OCNGS Technical Specification Proposed Technical Specification Marked-Up Page The page included in this attachment is:

4.2-1

REACTIVITY CONTROL Applies to surveillance requirements reactivity CDrltrC)1.

Objective: To verify the capability for controlling reactivity.

Specification:

A. SOM shall be verified:

1. Prior to each CORE ALTERATION, and

2. Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the first criticality following any CORE ALTERATION.

B. The control rod drive housing support system shall be inspected after reassembly.

C. The maximum scram insertion time of the control rods shall be demonstrated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:

1. For all control rods prior to THERMAL POW ER exceeding 40% power with reactor coolant pressure greater than 800 psig, following core alterations or after a reactor shutdown that is greater than 120 days.

2. For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "a" or "b" as follows:

a.1 Specifically affected individual control rods shall be scram time tested with the reactor depressurized and the scram insertion time from the fully withdrawn position to 90% insertion shall not exceed 2.2 seconds, and a.2 Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure prior to exceeding 40% power.

b. Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure.

3. On a frequency of less than or equal to once per 180 days of cumulative power operation, for at least 20 control rods, on a rotating basis, with ~ I;")

reactor coolant pressure greater than 800 psig. ~ Oltyy O. Each~rtiAIIY O[tul};;Jwithdrawn control rod shall be exercised at least once~

~ This test shall be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation is continuing with two or more inoperable control rods or in the event power operation is continuing with one fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out. The surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable rods has been reduced to less than two and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.

OYSTER CREEK 4.2-1 Amendment No: 178, 198, a49,266