PLA-5563, Response to Request for Additional Information for Proposed Amendment No. 211 to Unit 2, MCPR Safety Limits and Reference Changes

Response to Request for Additional Information for Proposed Amendment No. 211 to Unit 2, MCPR Safety Limits and Reference Changes

ML023610427
Person / Time
Site:	Susquehanna
Issue date:	12/18/2002
From:	Shriver B Susquehanna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-5563, TAC MB5610
Download: ML023610427 (7)

Text

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i PPL Susquehanna, LLC *' '* b I !I a '

Bryce L. Shriver 769 Salem Boulevard Senior Vice President and Berwick, PA 18603 Chief Nuclear Officer Tel. 570 542.3120 Fax 570.542.1504 P1 blshriver@ pplweb corn liabar.. TMwI,.

DEC 18 2002 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Mail Stop OP1-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION FOR REQUEST FOR ADDITIONAL INFORMATION 2

PROPOSED AMENDMENT NO. 211 TO UNIT LICENSE NPF-22: MCPR SAFETY LIMITS AND REFERENCE CHANGES Docket No. 50-388 PLA-5563 "ProposedAmendment No. 211 to Unit 2

Reference:

1) PLA-5467, R. L. Anderson (PPL) to USNRC, ", dated July 17, 2002.

License NPF-22: MCPR Safety Limits and Reference Changes to Proposed Amendment

2) PLA-5520, B. L. Shriver (PPL)to USNRC, "Supplement and Reference Changes",

No. 211 to Unit 2 License NPF-22: MCPR Safety Limits dated October 30, 2002.

Electric Station, Unit 2 - Request

3) USNRC to B. L. Shriver (PPL), "SusquehannaSteam Ratio Safety Limits and for Additional Information (RAI) - Minimum CriticalPower 9, 2002.

Reference Changes (TAC No. MB5610), dated December Request for Additional The purpose of this letter is to provide a response to the NRC Proposed Amendment Information (RAI), (Reference 3). The RAI refers to PPL's Safety Limits and Reference No. 211 to Unit 2 License No. NPF-22 relating to MCPR changes to the Unit 2 Changes. Specifically, Proposed Amendment No. 211 identified (TS) Section 2.1.1.2, Cycle 12 (U2C12) MCPR Safety Limits in Technical Specification to the Design Features in changes to the references in TS Section 5.6.5.b, and a change to TS Section 5.6.5 TS Section 4.2.1, (Reference 1). Supplemental information related was provided by Reference 2 in late October 2002.

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Document Control Desk PLA-5563 to this letter contains the RAI questions and responses as prepared by PPL.

These questions and responses were previously discussed via telecon between NRC and PPL on November 20, 2002.

Any questions regarding this additional information should be directed to Mr. Duane L. Filchner at (610) 774-7819.

Sincerely, B.L. Shriver Attachments: Affidavit Attachment 1 - Response to NRC's Request for Additional Information copy: NRC Region I Mr. D. J. Allard, PA DEP Mr. T. G. Colburn, NRC Sr. Project Manager Mr. S. Hansell, NRC Sr. Resident Inspector Mr. R. Janati, DEP/BRP

9-1.

BEFORE THE BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of PPL Susquehanna, LLC: Docket No. 50-388 REQUEST FOR ADDITIONAL INFORMATIONREGARDING PROPOSED AMENDMENT NO. 211 TO UNIT 2 LICENSE NPF-22: MCPR SAFETY LIMITS AND REFERENCE CHANGES Licensee, PPL Susquehanna, LLC, hereby files a revision to its Facility Operating License No. NPF-22 dated March 23, 1984.

This amendment involves a revision to the Susquehanna SES Unit 2 Technical Specifications.

PPL Susquehanna, LLC thi /-.- da o-7202 BAL. Snirver Senior Vice-President and Chief Nuclear Officer Sworn to and subscribed before me Notarial Seal Nancy J. Lannen, Notary Public Allentown, Lehigh County My Commission Expires June 14, 2004

• L otary Public

r Attachment I to PLA-5563 Response to NRC's Request for Additional Information C

Attachment 1 to PLA-5563 Page 1 of 3 Attachment 1 - Response to RAI Request for Additional Information for Proposed Amendment No. 211 to Unit 2 License NPF-22:

MCPR Safety Limits and Reference Changes NRC Ouestion 1:

Identify any differences between the reference core loading pattern for the analysis and the real core loading pattern. Provide the procedure of how to deal with this difference if it does occur.

PPL Response:

There is no recent experience where the Susquehanna real core loading pattern differed from the reference core loading pattern as contained in the FSAR. The current U2C 11 core loading pattern is in Section 4 of the Susquehanna Final Safety Analysis Report.

Should a core loading pattern different from the reference be deerhed warranted, PPL would assess the change in accordance with the PPL procedure for 10 CFR 50.59 to determine if the change requires NRC review and approval prior to implementation.

For Unit 2 Cycle 11 (U2Cl 1), Enclosure E to PLA-5169, (Referehice 1.1 below), specifies that U2C1 1 has 300 fresh ATRIUMTM-1O fuel assemblies, 280 once-burned ATRIUMTM 10 fuel assemblies, and 184 twice-burned ATRIUMTM-l0 fuel assemblies.

For Unit 2 Cycle 12 (U2C12), Attachment 5 to PLA-5467, (Reference 1.2 below),

specifies that U2C12 contains 284 fresh ATRIUMTM-10 fuel assemblies, 300 once burned ATRIUMTM-10 fuel assemblies, and 180 twice-burned ATRIUMTM-l0 fuel assemblies.

In the industry, core loading pattern changes have been necessitated as a result of fuel failures. Susquehanna has not experienced a fuel failure since 1992.

References:

1.1 PLA-5169, R. G. Byram (PPL) to USNRC, "Proposed Amendment No. 194 to License NPF-22: MCPR Safety Limits", dated March 20, 2000.

Attachment 1 to PLA-5563 Page 2 of 3 1.2 PLA-5467, B. L. Shriver (PPL) to USNRC, "Proposed Amendment No. 211 to Unit 2.

License NPF-22: MCPR Safety Limits and Reference Changes", dated July 17, 2002.

NRC Ouestion 2:

On page 1, attachment 1 of your submittal, you state that NRC approval of the previously used computer code ANFB-10 critical power correlation required a factor of 2 to be applied to the number of pins calculated to be in boiling transition for the Safety Limit calculation. Provide a detailed basis of this statement and identify the impact on the Safety Limit calculation.

PPL Response:

The basis for the statement with respect to the need for a factor of 2 on number of pins in boiling transition required for the ANFB correlation is found in the response to an RAI given in Reference 2.1, Supplement 2, page 7.

The previously used NRC approved ANFB critical power correlation had a mean bias in the predicted to measured ratio of critical power that was slightly greater than 1.0 (i.e., 1.003) (Reference 2.1 Supplement 2, page 7 below). This ratio of 1.003 means that the ANFB correlation on average would result in a predicted critical power slightly higher than the actual critical power (i.e., nonconservative). To compensate for this slight non conservatism in the ANFB correlation mean, a factor of 2 was applied to the number of pins in boiling transition when calculating the MCPR Safety Limit.

For the currently used NRC approved ANFB-10 critical power correlation, the mean bias in the ratio of predicted to measured critical power is less than 1.0 (i.e., 0.9985),

(Reference 2.2, Rev. 0 Page 4-2 below). Because the correlation inean is conservative, there is no need to apply any additional conservative factor. The impact on the number of pins in boiling transition is that there is no factor of 2 applied to the number of rods in boiling transition with the ANFB-10 based calculation. The resulting ANFB-10 based MCPR Safety Limit still assures that at least 99.9% of the fuel rods are expected to avoid boiling transition during normal operation and anticipated operatibnal occurrences.

Attachment I to PLA-5563 Page 3 of 3

References:

2.1 ANF-524 (P)(A), Revision 2 and Supplements, "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors",

November 1990.

2.2 EMF-1997 (P)(A), Revision 0, "ANFB-10 Critical Power Correlation," and EMF-1997 (P)(A) Supplement 1 Revision 0, "ANFB-10 Critical Power Correlation:

High Local Peaking Results", July 1998.