RA-11-005, Biennial 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports - January 1, 2009 Though December 31, 2010

Biennial 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports - January 1, 2009 Though December 31, 2010

ML11189A062
Person / Time
Site:	Oyster Creek
Issue date:	06/24/2011
From:	Massaro M Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-11-005
Download: ML11189A062 (6)

Text

Oyster Creek Generating Station www.exeloncorp.com Exelkn Route 9 South PO Box 388 Forked River, NJ o8731 10 CFR 50.59 10 CF 72.48 RA-1 1-005 June 24,2011 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Biennial 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports -

January 1, 2009 though December 31, 2010 Enclosed are the Oyster Creek Nuclear Generating Station 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports for regulatory commitments changed during the period of January 1, 2009 through December 31, 2010.

There are no regulatory commitments contained in this submittal.

Please contact Jeff Chrisley at (609) 971-4469 if any further information or assistance is needed.

Sincerely, Michael J. Massaro Vice President Oyster Creek Nuclear Generating Station Enclosure cc: Administrator, USNRC Region I G. Edward Miller, USNRC Senior Project Manager, Oyster Creek J. Kuip,. USNRC Senior Resident Inspector, Oyster Creek

Enclosure - RA-1 1-005 Page 1 of 5 Exelon Generation Company, LLC Oyster Creek Nuclear Generating Station Enclosure to RA-1 1-005 Docket No. 50-219 72-15 2009 - 2010 Biennial 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports These summary reports are issued pursuant to reporting requirements for Oyster Creek Nuclear Generating Station (OCNGS). These reports address tests, experiments, and changes to the facility and procedures as they are described in the Final Safety Analysis Report for the OCNGS station and the Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System (TN-61 BT Spent Fuel Cask at OCNGS).

These reports summarize the three tests, experiments, and changes that were implemented between January 1, 2009 and December 31, 2010 under 10 CFR 50.59.

There were no tests, experiments, or changes implemented by OCNGS under 10 CFR 72.48.

Enclosure - RA-1 1-005 Page 2 of 5 Evaluation Number: OC-2009-E-001 PORC Review Meeting No. (Date): 2009-02 (9/29/2009)

ActivitylDocument No.:

Title:

ESW and SW Cross Connect Description of Activity:

A cross connect was installed between the Service Water (SW) and Emergency Service Water (ESW) piping downstream of the Reactor Building Closed Cooling Water Heat Exchangers.

Reason for Activity:

Emergent repairs were required on the SW piping downstream of the Reactor Building Closed Cooling Water heat exchangers.

Effect of Activity:

The cross-connect from the SW to the ESW will only be used at times when both ESW systems are not required to be operable. The UFSAR design basis function of the SW and of the ESW and Containment Spray System are not adversely affected when they are required to be in operable status. The SW to ESW System 1 cross-tie will only be used when ESW is not required to be operable after it is verified that the flow rate, which is available through the cross-tie, is adequate for the removal of plant heat loads cooled by the Reactor Building Closed Cooling Water System.

Summary of Conclusion for the Activity's 50.59 Review:

There are no adverse procedure changes as a result of this activity. This activity does not involve a test or experiment. There are no changes required to be made to Technical Specifications or Operating License.

Evaluation Number: OC-201 0-E-001 PORC Review Meeting No. (Date): W1 0-03 (11/17/2010)

ActivitylDocument No.: Design ChangelECR OC 09-00866

Title:

TRACGO4 version 4.2.60.3 Description of Activity:

This activity addresses the acceptability of applying version 4.2.60.3 of TRACGO4P to determine Average Power Range Monitor (APRM) stability protection settings as documented in the Core Operating Limits Report (COLR). Version 4.2.60.3 of TRACG04P is an upgraded version of the NRC approved TRACG02A program originally developed and licensed to determine stability protection setpoints. Version 4.2.60.3 of TRACG04P had not been generically approved by the NRC. Therefore, applying TRACG04P version 4.2.60.3 reload analysis constituted a potential change in methodology and was evaluated in accordance with 10 CFR 50.59.

Enclosure - RA-1 1-005 Page 3 of 5 Reason for Activity:

GEH has implemented a PC-based version of the TRACG04 program (TRACG04P version 4.2.57.11) and recently upgraded the program (version 4.2.60.3) to address a number of programming issues identified since its initial release. The use of TRACG04P version 4.2.60.3 for OCNGS in determining stability protection settings constitutes a potential change in methodology. The upgraded version of the code was developed under the GEH NRC-approved Quality Assurance Program. However, since TRACG04P version 4.2.60.3 had not been reviewed and approved by NRC and GEH is not a license holder, the change was evaluated under 50.59 for use at OCNGS.

Effect of Activity:

OCNGS has implemented stability long-term solution Option II as stated in section 4.3.2.7.2 of the UFSAR. The TRACG thermal-hydraulic code supports the determination of the APRM stability setpoints protection settings in Section 2.3 of the OCNGS Technical Specifications. Evaluation of the applicability of TRACGO4P Version 4.2.60.3 ensures that the cycle specific stability setpoints determined using the newer version of TRACG are within the requirements established by the NRC's review and approval of TRACG for this .purpose.

Summary of Conclusion for the Activity's 50.59 Review:

GEH benchmarking analysis confirms that Version 4.2.60.3 of TRACGO4P produces results that are essentially the same or slightly more conservative (more limiting) than those produced by TRACGO2A. Therefore, in accordance with 10 CFR 50.59(a)(2)(i),

the use of TRACGO4P Version 4.2.60.3 does not constitute a departure from a method of evaluation described in the UFSAR, and TRACGO4P can be used to support the determination of cycle specific APRM stability protection setpoints without prior NRC approval.

Evaluation Number: OC-201 0-E-002 PORC Review Meeting No. (Date): 2010-09 (11/22/2010)

ActivitylDocument No.: Design Change/ECR OC 09-00788 R1

Title:

GNF2 Alternate Source Term Analysis Description of Activity:

The proposed activity is a configuration change (ECR OC 09-00788 Revision 001) that supports the implementation of GNF-2 as a new fuel type at OCNGS. This evaluation will specifically address the impact of the new GNF-2 fuel on the design analyses that determine offsite and control room radiological consequences of the following accident conditions evaluated in UFSAR Chapter 15:

• Fuel Handling Accident (FHA)

• Control Rod Drop Accident (CRDA)

• Main Steam Line Break (MSLB)

• Loss of Coolant Accident (LOCA)

Enclosure - RA-1 1-005 Page 4 of 5 The proposed activity uses the containment leakage and Main Steam Isolation Valve (MSIV) leakage information specified in the OCNGS Technical Specifications and the Alternative Source Term (AST) methodology and Total Effective Dose Equivalent (TEDE) dose criteria approved in Amendment No. 262 to the OCNGS Operating License. In addition to the revised GNF-2 source terms, the proposed activity includes the following changes to LOCA analysis design inputs of the RADTRAD analysis code to be consistent with RG 1.183 and the Polestar analysis supporting Amendment 262:

" Revised leakage rate to credit time delay for leakage to travel from outboard MSIV to Turbine Stop Valve (TSV) before being released to the environment (8.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> with inboard MSIV open and 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> with both MSIVs closed)

" Revised time at which primary leakage rate can be reduced by 50% to after first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> rather than after first 15.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

• Revised to delete credit for aerosol removal in the Wetwell in evaluating containment leakage and bypass leakage pathways

" Revised model used to establish engineered safety feature leakage to be consistent with RG 1.183 Appendix A The proposed activity includes changes to the UFSAR (Change #01126203-04) that incorporate the results of the GNF-2 analyses and the application of the AST methodology and the TEDE criteria of 10 CFR 50.67. As documented in IR 1115400, the UFSAR was not updated following approval of Amendment 262 to the OCNGS Operating License in April of 2007.

The proposed activity also includes a change to the Technical Specification Bases (Section 4.5) that revises the maximum Exclusion Area Boundary and Control Room LOCA dose based on the results of the GNF-2 analyses.

Reason for Activity:

The introduction of a new fuel type affects the source term used in the design analyses that determine Exclusion Area Boundary, Low Population Zone, and Control Room LOCA dose for accident conditions evaluated in UFSAR Chapter 15.

Effect of Activity:

The proposed activity establishes new offsite and control room radiological consequences for FHA, CRDA, MSLB and LOCA; however, the revised consequences do not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR. In addition, the proposed activity will not result in a design basis limit for a fission product barrier being altered or exceeded. The evaluation methodology is consistent with the AST methodology approved in Operating License Amendment 262 to the OCNGS Operating License, .RG 1.183, and the associated NRC Safety Evaluation Report approving Amendment 262.

Summary of Conclusion for the Activity's 50.59 Review:

The proposed change to design analysis input parameters used in determining offsite and control room radiological consequences does not result in operation of equipment outside the design functions as currently described in the UFSAR. The new GNF-2 fuel type will perform the same functions within the same operational limits as the current fuel

Enclosure - RA-1 1-005 Page 5 of 5 types in use at OCNGS. The malfunctions and non-radiological accidents currently analyzed in the UFSAR are not affected by the revised design analysis input parameters. There are no new system interfaces created by the proposed activity and no physical changes are made to the environment or release paths evaluated in the design analyses. As a result, the proposed activity does not increase the likelihood of a malfunction of equipment important to safety, does not create the possibility for an accident or malfunction of equipment important to safety of a different type than previously analyzed in the UFSAR, and does not increase the frequency of accidents previously evaluated in the UFSAR.

Therefore, in accordance with 10 CFR 50.59(a)(2)(i), the activity can be implemented without prior NRC review and approval.