ML13137A103

2013/03/28 NRR E-mail Capture - Monticello Nuclear Generating Plant - Draft Requests for Additional Information (Srxb) Review of Extended Power Uprate (TAC No. MD9990)

ML13137A103
Person / Time
Site:	Monticello
Issue date:	03/28/2013
From:	Beltz T A Plant Licensing Branch III
To:	Fields J Northern States Power Co
References
TAC MD9990
Download: ML13137A103 (4)

Text

1NRR-PMDAPEm ResourceFrom:Beltz, TerrySent:Thursday, March 28, 2013 9:58 AMTo:'Fields, John S.'Cc:'gene.eckholt@xenuclear.com'; Jackson, Christopher; Carlson, Robert; Parks, BenjaminSubject:Monticello Nuclear Generating Plant - Draft Requests for Additional Information (SRXB) re: Review of Extended Power Uprate (TAC No. MD9990)Attachments:Monticello - Draft Requests for Additional Information (SRXB) Associated with Gap Analysis Review .docx Dear Mr. Fields:

The U.S. Nuclear Regulatory Commission (NRC) staff in the Reactor Systems Branch (SRXB) of the Office of Nuclear Reactor Regulation is currently supporting the review of the Extended Power Uprate (EPU) license amendment request for the Monticello Nuclear Generating Plant. The staff has reviewed your January 21, 2013, supplemental response associated with the EPU Gap Analysis (ADAMS Accession Package No.

ML130390220) and determined that additional information is required to complete its review. Draft requests for additional information (RAI) are attached. You may accept these draft RIAs as formal Requests for Additional Information and respond to the questions by May 3, 2013. Alternatively, you may request to discuss the contents of these RAIs with the NRC staff in a conference call, including any change to the proposed response date.

Please let me know if you have any questions or concerns.

Sincerely, Terry A. Beltz, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation (301) 415-3049 Terry.Beltz@nrc.gov Hearing Identifier: NRR_PMDA Email Number: 699 Mail Envelope Properties (87B1F1BDFE5A554CA9DC5EAA75EB6D0DCDEFEDAA82)

Subject:

Monticello Nuclear Generating Plant - Draft Requests for Additional Information (SRXB) re: Review of Extended Power Uprate (TAC No. MD9990) Sent Date: 3/28/2013 9:57:51 AM Received Date: 3/28/2013 9:57:00 AM From: Beltz, Terry Created By: Terry.Beltz@nrc.gov Recipients: "'gene.eckholt@xenuclear.com'" <gene.eckholt@xenuclear.com> Tracking Status: None "Jackson, Christopher" <Christopher.Jackson@nrc.gov> Tracking Status: None "Carlson, Robert" <Robert.Carlson@nrc.gov>

Tracking Status: None "Parks, Benjamin" <Benjamin.Parks@nrc.gov> Tracking Status: None

"'Fields, John S.'" <John.Fields@xenuclear.com> Tracking Status: None Post Office: HQCLSTR01.nrc.gov Files Size Date & Time MESSAGE 1186 3/28/2013 9:57:00 AM Monticello - Draft Requests for Additional Information (SRXB) Associated with Gap Analysis Review .docx 19529 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: Recipients Received:

REQUESTS FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REACTOR SYSTEMS BRANCH EXTENDED POWER UPRATE REVIEW NORTHERN STATES POWER COMPANY - MINNESOTA (NSPM) MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 TAC NO. MD9990 1. Page 16 of Enclosure 1 to the January 21, 2013, letter (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13039A200) discusses Item 8 of the Extended Power Uprate (EPU) Gap Analysis, concerning emergency core cooling system (ECCS) pump flow rates. The response refers to additional correspondence (ADAMS Accession Nos. ML12276A057 and ML12276A057) which, based on cursory review, appear to indicate that some assumptions and analyses credit revised ECCS pump flow rates that remain bounded by the SAFER ECCS evaluation.

Please confirm that the SAFER ECCS evaluation includes ECCS pump flow rates that are bounding of these revised ECCS pump flow rate assumptions. 2. Page 20 of Enclosure 1 to the January 21, 2013, letter discusses Item 10 of the EPU Gap Analysis, concerning the effects of a final feedwater (FW) temperature change. It states, "[General Electric-Hitachi (GEH)] performed a study and determined that the impact of the FW temperature change on anticipated operational occurrences (AOOs) was negligible." Please describe how the study was performed and provide additional information regarding the basis for this determination. 3. Page 20 of Enclosure 1 to the January 21, 2013, letter discusses the effects of a final FW temperature change. It states, "GEH further concluded that sufficient margin remains in the peak dome pressure safety limit and ASME upset condition limit when accounting for this small FW temperature change." Describe how this conclusion was reached. Explain how much margin is required to offset the effects of the final FW temperature change, how the amount of margin remaining in these limits was determined, and how MNGP will ensure that adequate margin is maintained in cycle-specific safety analyses.

- 2 - 4. Page 20 of Enclosure 1 to the January 21, 2013, letter discusses the effects of a final FW temperature change. The applicable section describes and evaluation of the design basis accident (DBA) - loss of coolant accident (LOCA) containment response. The section does not describe the effects that the final FW temperature change could have on the DBA-LOCA ECCS evaluation. Please explain how the EPU ECCS evaluation accounts for the final FW temperature change. 5. Page 40 of Enclosure 1 to the January 21, 2013, letter discusses Item 15 of the EPU Gap Analysis, concerning a change in the turbine bypass valve capacity value, which apparently amounted to a slight reduction, i.e., from 11.6% to 11.5%. The section states that "the evaluation of plant transients is performed on a cyclic basis for MNGP and has been completed for EPU core design using a value of 11.5% for the evaluation of transients- the results of this- evaluation are available in the MNGP cycle 26 supplemental reload licensing report-"

Please address the effects of this change with respect to the limiting ATWS overpressure events. 6. Page 49 of Enclosure 1 to the January 21, 2013, letter discusses Item 21 of the EPU Gap Analysis, concerning reported errors and changes in the ECCS evaluation. The section states the following: Notification 2012-01 [reported to the NRC by letter dated December 26, 2012 (ML12363A073)], which is related to implementation of the GEH PRIME thermal-mechanical model, is not considered an Evaluation Model Error, but rather a Change. Note that, as the implementation of PRIME-based analytical methods addresses an issue where the prior, GESTR-M-based ECCS evaluation model calculated peak cladding temperatures that were less conservative than previously understood due to the inability of GESTR-M to account for the degradation of fuel thermal conductivity with increasing exposure, the NRC staff disagrees with this characterization. Refer to Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation," for a similar characterization offered by a fuel vendor, with which the NRC staff also disagreed. The NRC staff does not, as a matter of general practice, approve ECCS evaluations that are known to contain errors.

Please submit an updated ECCS evaluation that incorporates the PRIME-based model.