ML100820298

Requests for Additional Information, License Amendment Request to Adopt the Alternative Source Term Methodology

ML100820298
Person / Time
Site:	Prairie Island
Issue date:	03/26/2010
From:	Beltz T Plant Licensing Branch III
To:	Schimmel M Northern States Power Co
beltz T, NRR/DORL/LPL3-1, 301-415-3049
References
TAC ME2609, TAC ME2610
Download: ML100820298 (6)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 26, 2010 Mr. Mark A. Schimmel Site Vice President Prairie Island Nuclear Generating Plant Northern States Power - Minnesota 1717 Wakonade Drive East Welch, MN 55089 SUB~IECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 REQUESTS FOR ADDITIONAL INFORMATION RE: LICENSE AMENDMENT REQUEST TO ADOPT THE ALTERNATIVE SOURCE TERM METHODOLOGY (TAC NOS. ME2609 AND ME2610)

Dear Mr. Schimmel:

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 27,2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML093160583), Northern States Power Company, a Minnesota corporation (NSPM), submitted a license amendment request for the Prairie Island Nuclear Generating Plant, Units 1 and 2.

The proposed amendment would adopt the Alternative Source Term (AST) methodology, in addition to technical specification changes supported by the AST design basis accident radiological consequences analysis. The proposed amendment would also incorporate Technical Specification Task Force (TSTF)-490, "Deletion of E-Bar Definition and Revision to RCS [reactor coolant system] Specific Activity Tech Spec," Revision O.

To complete their technical review, the NRC staff provided draft requests for additional information (RAls) to NSPM (ADAMS Accession Nos. ML100620213, ML100680414, and ML100750041). The draft RAls were discussed with your staff to resolve outstanding concerns or provide additional clarity regarding the requested information.

The finalized RAls are being issued as an enclosure to this letter. As agreed upon with Ms. Amy Hazelhoff, please respond to the enclosed RAls no later than April 30, 2010.

M. A. Schimmel

- 2 If you have any questions, please contact me at (301) 415-3049.

Sincerely,

~-~-

Terr A.. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosure:

As stated cc w/encl: Distribution via ListServ

REQUESTS FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST (LAR) REGARDING ALTERNATIVE SOURCE TERM (AST)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306 TAC NOS. ME2609 AND ME2610 ELECTRICAL ENGINEERING BRANCH (EEEB)

EEEB RAI1 Provide the changes in the emergency diesel generator (EDG) loading due to this LAR.

Provide the EDG loads for Units 1 and 2 pre-and post-AST, and confirm that adequate margin exists post-AST. Confirm that EDG testing envelopes the loading requirements due to this LAR.

EEEB RAJ 2 Confirm that no credit has been taken for non-safety related system(s) due to this LAR.

If yes, then a) describe the impact on the EDG loading due to these systems, b) describe how electrical and physical separation, and single failure criteria, have been met, and c) describe how the operators will be notified (e.g., control room annunciators) in the event that these systems become inoperable.

EEEB RAI3 Confirm that there are no changes to the environmental qualification program due to this LAR.

MECHANICAL AND CIVIL ENGINEERING BRANCH (EMCB)

EMCB RAI1 Page 9 of the enclosure in Reference 1 indicates that credit will be taken for isolation dampers, which make up a portion of the Auxiliary Building Normal Ventilation System, as part of the proposed AST implementation. Various justifications are provided for crediting these non-safety related dampers as part of the proposed implementation. However, no justification is provided regarding the structural and seismic ruggedness of this equipment. Appendix A to 10 CFR Part 100 requires that structures, systems, and components necessary to assure the capability of the plant to mitigate the consequences of accidents, which could result in exposures comparable to the guideline exposures provided in 10 CFR Part 100, be designed to remain functional during and after a safe shutdown earthquake.

Enclosure

- 2 Please discuss the methodologies used to demonstrate the seismic ruggedness and/or seismic qualification of the aforementioned dampers. Additionally, please provide the references which provide the regulatory acceptance bases of these methodologies.

REACTOR SYSTEMS BRANCH (SRXB)

The following requests for additional information are associated with the NRC staff review of the steam generator tube rupture (SGTR) overfill analysis in support of the AST LAR application, as discussed in the Enclosure to Reference 1, Pages 106 and 107.

Please provide the following information for the NRC staff to continue its review:

SRXB 1 Discuss the methods used for the SGTR overfill analysis. If the methods were previously approved by NRC, list the NRC safety evaluation reports approving the methods. If the methods were not reviewed and approved by NRC, address acceptability of the methods.

The information should include a description and justification of reactor coolant system (RCS) models with safety injection simulation, models for determination of the primary-to-secondary break and steam relief flow rates, and steam generator (SG) water level model accounting for the effects of bubble formation during depressurization on the SG water level for a SGTR event.

SRXB 2 Provide a list the nominal values with measurement uncertainties and the corresponding values used in the SGTR overfill analysis for the following applicable plant parameters:

Initial RCS pressure Initial SG water inventory Safety injection actuation pressure setpoint Safety injection flow versus RCS pressure Safety injection system pump delay time SG relief valve pressure setpoint Auxiliary feedwater actuation setpoint and delay time Auxiliary feedwater flow rate per SG Auxiliary feedwater temperature Time of loss of offsite power Delay times for reactor trip and turbine trip Decay heat model and initial value in percentage of the rated power level Discuss the effects of an increase or decrease in the value for each of the above plant parameters on the SG water level calculations during a SGTR and address the adequacy of the values used in the SGTR overfill analysis in minimizing the margin to SG overfill.

- 3 SRXB 3 List operator action times for the following applicable operator actions as determined by the plant simulator in accordance with the Emergency Operating Procedure E-3 for a SGTR:

Identify and isolate the rupture SG Initiate RCS cooldown Initiate RCS depressurization Terminate safety injection flow Establish charging flow Establish RCS letdown Reopen pressurizer power-operated relief valve Discuss the operator actions credited in the SGTR overfill analysis and provide a sequence of events for the SGTR including the above operator action times, and calculated times for the RCS cooldown, RCS depressurization and equalization of RCS and ruptured SG pressure. The information should show that the operator actions and their associated times assumed in the analysis were identical with that determined by the plant simulator.

SRXB4 List the single failure events considered in the SGTR overfill analysis and identify the worst single failure used in the analysis that resulted in a minimum margin to the SG overfill.

SRXB 5 Provide the results of the SGTR overfill analysis for the following applicable plant parameters:

Pressurizer pressure versus time Secondary pressures and SG water volumes versus time for both intact and rupture SGs Total primary to secondary break flow and total integrate primary to secondary break flow versus time SG relief flow and integrated SG relief flow versus time for both intact and rupture SGs The results indicate the following: the calculated RCS break and SG relief flowrates are consistent with the primary and secondary pressures; there is no unexplainable thermal hydraulic phenomenon; the RCS pressure and the rupture SG secondary pressure are equal; and, the SG does not overfill with water.

References:

1. Letter from Mark Schimmel, Xcel Energy, to the U.S. Nuclear Regulatory Commission Document Control Desk, Re: Prairie Island Nuclear Generating Plant, Units 1 and 2, Dockets 50-282 and 50-306, License Nos. DPR-42 and DPR-60, "License Amendment Request to Adopt Alternative Source Term Methodology," dated October 27, 2009 (ADAMS Accession No. ML093160583).

ML100820298

• concurrence via memo

• • concurrence via e-mail OFFICE LPL3-1/PM LPL3-1/LA EEEB/BC*

EMCB/BC*

NAME IrBeitz BTuliy GWilson MKhanna DATE 03/26/10

~3/25/10 03/04/10 03/04/10 OFFICE SRXB/BC **

LPL3-1/BC LPL3-1/PM NAME GCranston RPascarelli TBeitz DATE 02/24/10 03/26/10 03/29/10