AEP-NRC-2015-05, Correction of Technical Specification Typographical Error

Correction of Technical Specification Typographical Error

ML15035A065
Person / Time
Site:	Cook
Issue date:	02/02/2015
From:	Scarpello M Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2015-05, TAC ME9560
Download: ML15035A065 (8)

Text

INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWMR One Cook Place Bridgman, MI 49106 A unit ofAmerican ElectricPower India naMichiganPower.com February 2, 2015 AEP-NRC-2015-05 10 CFR 50.4 Docket No.: 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC, 20555-0001 Donald C. Cook Nuclear Plant Unit 2 Correction of Technical Specification Typographical Error

References:

1. Letter from J. P. Gebbie, Indiana Michigan Power (I&M), to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Application to Revise Technical Specifications to Adopt Technical Specification Task Force (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process," dated September 12, 2012, Agencywide Documents Access Management System (ADAMS) Accession Number ML12268A047.

2. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments Re: Adoption of TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidate Line Item Improvement Process (TAC NOS. ME9560 and ME9561)," dated March 22, 2013, ADAMS Accession Number ML13046A114.

3. Proposed Guidance for Correction of Technical Specification Typographical Errors, SECY-96-238, dated November 19, 1996.

This letter requests the U. S. Nuclear Regulatory Commission's (NRC) approval of a correction to typographical errors in Donald C. Cook Nuclear Plant (CNP) Unit 2 Technical Specifications (TS) which were inadvertently introduced during License Amendment 304 (Reference 2).

These typographical errors were not addressed in the notice to the public, nor reviewed by the NRC, and fall within the scope of guidance provided in SECY-96-238 (Reference 3) for corrections. describes the typographical errors and corrections. Enclosure 2 provides a mark-up of the TS page. Enclosure 3 provides a clean copy of the TS page. Indiana Michigan Power will

U. S. Nuclear Regulatory Commission AEP-NRC-2015-05 Page 2 issue the change notification to all holders of controlled copies of CNP TSs upon receipt of NRC approval of these corrections.

This letter contains no new commitments. If you have any questions or require additional information, please contact me at (269) 466-2649.

Sincerely, Michael K. Scarpello Regulatory Affairs Manager DMB/kmh

Enclosures:

1. Description of Technical Specification Typographical Errors

2. Mark-up Technical Specification Page Correcting the Typographical Errors

3. Clean Copy of Technical Specification Page Correcting the Typographical Errors c: M. L. Chawla, NRC Washington DC J. T. King, MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III A. J. Williamson, AEP Ft. Wayne, w/o enclosure

ENCLOSURE I TO AEP-NRC-2015-05 Description of Technical Specification Typographical Errors A. Requested Action Consistent with the information contained in SECY-96-238, Proposed Guidance for Correction of Technical Specification Typographical Errors, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 2, is requesting correction for two inadvertent typographical errors that were introduced into the CNP Technical Specifications (TS).

B. Typographical Error Two typographical errors were introduced into the CNP TSs during implementation of Unit 2 License Amendment 304 (Reference 2). The specification of the errors and proposed corrections are described below:

Unit 2 Page 5.6-4 The references for "5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)"

lists a number "9. 5.6 Reporting Requirements." This reference should not be here as it is the title of the section that is listed at the top of the page.

This error was not introduced in the original License Amendment Request (LAR)

Mark-up pages (Reference 1) that were submitted to the U. S. Nuclear Regulatory Commission (NRC) and was, therefore, not noticed to the public, nor approved by the NRC during review and approval of Amendment 304. This error was introduced on the clean pages issued after NRC review.

Unit 2 Page 5.6-4 The first sentence in Section 5.6.6 Post Accident Monitorinq Report states: "When a report is required by Condition B or H of LCO 3.3.3 ...." This should state "...by Condition B or G of LCO..."

This error was not introduced in the original LAR Mark-up pages (Reference 1) that were submitted to the NRC and was, therefore, not noticed to the public, nor approved by the NRC during review and approval of Amendment 304. This error was introduced on the clean pages issued for Amendment 304 after NRC review.

C. Corrections to the Affected TS Page SECY 96-238, (Reference 3) provides guidance to correct inadvertent typographical errors in the TS pages. Specifically:

"...if the licensee can demonstrate that the error was introduced inadvertently in a particular license amendment and that the erroneous change was not addressed in the notice to the public nor reviewed by the staff. Under these limited circumstances, the change that introduced the typographical error was not a proper amendment to the to AEP-NRC-2015-05 Page 2 license because it was neither addressed in the notice nor reviewed, and correction of the typographical error is not a "change" to the TS. Accordingly, the typographical error may be corrected by a letter to the licensee from the NRC staff, instead of an amendment to the license.

The above typographical errors were not noticed to the public, nor reviewed by the NRC as part of the application amendment process. Therefore, these errors may be corrected without a license amendment.

Accordingly, upon approval from the NRC, a corrected Unit 2 TS Page 5.6-4 will be issued to all holders of controlled TS.

D. References

1. Letter from J. P. Gebbie, Indiana Michigan Power (I&M), to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Application to Revise Technical Specifications to Adopt Technical Specification Task Force (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process," dated September 12, 2012, Agencywide Documents Access Management System (ADAMS) Accession Number ML12268A047.

2. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments Re: Adoption of TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidate Line Item Improvement Process (TAC NOS. ME9560 and ME9561)," dated March 22, 2013, ADAMS Accession Number ML13046A114.

3. Proposed Guidance for Correction of Technical Specification Typographical Errors, SECY-96-238, dated November 19, 1996.

ENCLOSURE 2 TO AEP-NRC-2015-05 DONALD C. COOK NUCLEAR PLANT UNIT 2 MARK-UP TECHNICAL SPECIFICATION PAGE CORRECTING THE TYPOGRAPHICAL ERRORS

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary); and

8. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOTM, July 2006 (Westinghouse Proprietary)

9. 5.6 Reporting Requiremnents

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoringq Report When a report is required by Condition B or H4@ of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each degradation,

f. The number and percentages of tubes plugged to date, and the effective plugging percentage in each generator, and Cook Nuclear Plant Unit 2 5.6-4 Amendment No. 269, 270, 2-7-9, 296, 2-97 302, 304

ENCLOSURE 3 TO AEP-NRC-2015-05 DONALD C. COOK NUCLEAR PLANT UNIT 2 CLEAN COPY OF TECHNICAL SPECIFICATION PAGE CORRECTING TYPOGRAPHICAL ERRORS

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary); and

8. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOTM, July 2006 (Westinghouse Proprietary)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. 'The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each degradation,

f. The number and percentages of tubes plugged to date, and the effective plugging percentage in each generator, and Cook Nuclear Plant Unit 2 5.6-4 Amendment No. 269, 2-7-0, 2-7-9, 296, 29-7 302, 304,