L-MT-10-049, Response to Request for Additional Information for the Monticello Proposed Maximum Extended Load Line Limit Analysis Plus (Mellla+) Amendment

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Response to Request for Additional Information for the Monticello Proposed Maximum Extended Load Line Limit Analysis Plus (Mellla+) Amendment
ML102790375
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 09/28/2010
From: O'Connor T
Xcel Energy, Northern States Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-10-049, TAC ME3145
Download: ML102790375 (10)
v • d • e


Text

WITHHOLD ENCLOSURES 1 AND 3 OF ATTCHMENT 2 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 XceI Energy Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 September 28, 2010 L-MT-10-049 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket 50-263 Renewed License No. DPR-22

Subject:

Response to Request for Additional Information (RAI) for the Monticello Proposed Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Amendment (TAC ME3145)

References:

1) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, "License Amendment Request: Maximum Extended Load Line Limit Analysis Plus," L-MT-1 0-003, dated January 21, 2010, TAC ME3145, Accession No. ML100280558.
2) Email from Peter Tam (NRC) to Lynne Gunderson, John S. Fields (Xcel),

"Monticello - Draft RAI from Reactor Systems Branch re. Proposed MELLLA+

Amendment (TAC ME3145)," July 13, 2010.

3) Email from Peter Tam (NRC) to Ralph Hayes and Lynne Gunderson (Xcel),

"Monticello RAI re. Proposed MELLLA+ Amendment (TAC ME3145)," August 13, 2010.

Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, requested in Reference 1 an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications to allow operation within the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain.

In References 2 and 3, the Reactor Systems Branch of the Nuclear Regulatory Commission (NRC) requested additional information for the review of the Reference 1 MELLLA+ submittal. provides the responses to requests for additional information (RAIs) 17, 18, 19, 25, and 26 (partial) from References 2 and 3, and refers to Attachment 2 for the responses to RAIs 1 through 16, 20 through 24, 26 and 27.

Document Control Desk Page 2 contains responses provided by GE Hitachi Nuclear Energy (GEH) and has four enclosures. Enclosure 1 provides the responses for RAIs 1 through 16, 20 through 24, 26 and 27 and contains GEH proprietary information. Enclosure 2 of Attachment 2 is a redacted non-proprietary version of Enclosure 1 with the GEH proprietary information removed, and is suitable for public disclosure. Enclosure 3 provides supplemental information for the responses for RAIs 6, 8, 10 and 21 in CD-ROM. form and contains GEH proprietary information. There is no redacted version of the CD-ROM.

NSPM requests that the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390(a)4, as authorized by 10 CFR 9.17(a)4. An affidavit supporting the request to withhold Enclosures 1 and 3 is provided in Enclosure 4 of Attachment 2. Any comments with respect to the affidavit should be directed to Edward D. Schrull, GE Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, NC 28401.

In Attachment 3, NSPM is providing a replacement page for Reference 1 to correct a typographical error in Reference 1, Enclosure 1, Table 1 (page 4 of 14). The error is in row 2 of the table, "SR 3.1.1.1.16" should be "SR 3.3.1.1.16," however the Technical Specification markups provided in Reference 1 were correctly marked to delete SR 3.3.1.1.16.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct.

Exec don: Septembe *f 2010 Tim O'Connor Site -President Mot ello Nuclear Generating Plant Nor e States Power Company-Minnesota Attachments (3) cc: Administrator, Region Ill, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC Minnesota Department of Commerce

ATTACHMENT I NSPM RESPONSES TO THE REQUESTS FOR ADDITIONAL INFORMATION FOR THE PROPOSED MELLLA+ AMENDMENT RAIs 17, 18, 19, 25, and 26 (partial)

L-MT-10-049 Page 1 of 7 NRC RAI 17 Section 9.3.1.2 of NEDC-33435P/Revl presents a sensitivity analysis for different water level control strategiesduring A TWS. Have the Emergency OperatingProcedures(EOPs)in Monticello been updated to reflect the lessons learned from these simulations?

Response

The discussion in the MELLLA+ Safety Analysis Report, NEDC-33435P, Revision 1, (Reference RAI-17-1), Section 9.3.1.2 evaluates the range of operating conditions covered by the existing Emergency Operating Procedures (EOP(s)).

The EOP level control procedure for an Anticipated Transient Without Scram (ATWS) event direct the operator to lower the water level until "powerdrops below 3.5% OR all SRVs stay closed and DWpressure stays below 1.84 psig OR level reaches -126" (TAF)." The level control procedure then directs the operator to establish a reactor pressure vessel (RPV) level control band between this value and -149". For details see copy of procedure C.5-2007, Failure to Scram, submitted in response to SRXB RAI No. 2.8.3-3 in Reference RAI-17-2.

A sensitivity study was performed to compare the plant response to a water level control strategy at top of active fuel (TAF) (-126") and TAF-2 ft (-149"). Procedure C.5-2007 directs the operators to maintain water level between approximately -126" and -149". The sensitivity study evaluates the effect of controlling water level in this range on the best-estimate TRACG ATWS analysis.

No EOP changes are required because the study demonstrates successful event mitigation with the existing EOP procedures. This study included evaluation of the plant response with and without reactor depressurization. The need for reactor depressurization is covered by the existing EOPs and may occur depending on initial conditions and event severity.

For further discussion of the need for EOP changes see the response to RAI 6 in Reference RAI-17-3.

References:

RAI-17-1 GEH, NEDC-33435P, Revision 1, "Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus," December 2009.

RAI-17-2 Letter L-MT-09-049 from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to NRC Document Control Desk, "Monticello Extended Power Uprate: Response to NRC Reactor Systems Review Branch and Nuclear Code and Performance Review Branch Request for Additional Information (RAI) dated March 23, 2009 and Nuclear Code and Performance Review Branch Request for Additional Information dated April 27, 2009 (TAC No. MD9990)," dated July 23, 2009 (ML092090219).

RAI-17-3 Letter L-MT-10-017 from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to NRC Document Control Desk, "Monticello MELLLA +: Supplemental Information Needed to Complete the Acceptance Review, (TAC ME3145)," dated March 4, 2010 (ML100710445).

L-MT-10-049 Page 2 of 7 NRC RAI 18 What are the net positive suction head (NPSH) requirements for critical equipment in Monticello? Provide an evaluation of NPSH requirements versus the predictedA TWS conditions.

Response

By letter dated October 1, 2009 (Reference RAI-1 8-1) from Eric Leeds of Office of Nuclear Reactor Regulation to T J O'Connor of NSPM, the NRC notified NSPM of a delay in the review schedule for Extended Power Uprate (EPU) until appropriate review criteria for containment accident pressure (CAP) credit becomes available. The review criteria have not been completed to date. As such this question cannot be responded to at this time for systems taking suction from the suppression pool.

In the MELLLA+ application (Reference RAI-18-2) NSPM committed to resolving the CAP issue for MELLLA+ in the same manner as it is resolved for the delayed EPU amendment.

References:

RAI-18-1 Monticello Nuclear Generating Plant -Revised Schedule for Review of Extended Power Uprate Amendment Application, (TAC NO. MD9990), dated October 1, 2009 (ML092600850).

RAI-18-2 License Amendment Request: Maximum Extended Load Line Limit Analysis Plus, dated January 21, 2010 (ML100280558)

L-MT-10-049 Page 3 of 7 NRC RAI 19 What is the status of the Monticello plant simulatorwith respect to MELLLA+? Provide a schedule of upgrades and validations that ensuresproper operatortraining priorto operation in the MELLLA+ domain.

Response

The Monticello plant simulator has not currently been modified to incorporate MELLLA+.

Simulator modifications to support operation under MELLLA+ conditions are scheduled to be installed and tested in the Spring 2011 refueling outage. However, this schedule may change as necessary, depending on the approval status of the MELLLA+ license amendment. The modification process and the operator training program require that operator training is completed prior to implementation of modifications in the plant. This ensures proper operator training occurs prior to operation in the MELLLA+ domain.

L-MT-1 0-049 Page 4 of 7 NRC RAI 25 NRC Generic Letter (GL) 88-16 provides guidance for technical specification changes for cycle-specific parameterlimits in the Core OperatingLimits Report (COLR), which requires that: (1) to identify which particularcycle-specific core operatinglimits listed in TS 5.6.3. a will be supported by the referenced approved methodologies listed in TS 5.6.3.b for calculatingthe cycle-specific operatinglimit listed in TS 5.6.3.a; and (2) to identify the supported approved methodologies by reportnumber, title, revision, date, and any supplements. TS 5. 6.3.b states that the COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). The proposed change to TS 5.6.3.b.4 follows the statement in TS 5.6.3.b and the GL 88-16 guidance. Please provide: (1) justification for not identifying the date, revision or supplement for TS 5.6.3.,b. I through TS 5.6.3.b. 3; (2) identification of which applicable cycle-specific parameterslisted in TS 5.6.3.a are supported by TS 5.6.3.b. I through TS 5.6.3.b.3; and (3) a COLR reportreflected in the attachment I of L-MT-1O-003 Markups to the Technical Specifications for MELLLA Plus.

Response

1. TS 5.6.3 was modified to remove unnecessary level of detail in TS Amendment Number 146 which was the Improved Technical Specifications (ITS) conversion (Reference RAI 01). The ITS conversion was approved by the NRC in the associated Safety Evaluation (SE). The SE specifically accepted the NSPM changes to TS 5.6.3.b as Type 3 changes, "Removing Procedural Details for Meeting TS Requirements or Reporting Requirements" in Table LA of the SE (ADAMS Accession Number ML061240266). These references are not affected by MELLLA+ and therefore have not been changed.
2. The following cycle-specific parameters are supported by TS 5.6.3.b.1, NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel"; (GESTAR):
  • .TS 5.6.3.a.4 Control Rod Block Instrumentation Allowable Value for the Table 3.3.2.1-1 Rod Block Monitor Functions 1.a, 1.b, and 1.c and associated Applicability RTP levels The Reactor Protection System Instrumentation Delta W Allowable Value for Table 3.3.1.1-1, Function 2.b, APRM Simulated Thermal Power- High, Note b, is a cycle specific parameter, however, Delta W only changes if there is a change to the thermal hydraulic configuration of the core and recirculation system. These changes could include, Increased Core Flow (ICF), power uprate, and/or physical changes to the recirculation piping, drive pumps or jet pumps. If any of these changes occur, then Delta W is evaluated with respect to that change. Delta W is only of concern for single loop operation (SLO) and SLO is not allowed in the MELLLA+ operating domain, therefore, Delta W does not affect MELLLA+.

L-MT-10-049 Page 5 of 7

3. In discussion with the NRC during a conference call held on August 2, 2010, NSPM stated that a COLR for a MELLLA+ core is not currently available. A non-MELLLA+ COLR is available and was requested by the NRC. The Monticello Cycle 25 COLR is available in ADAMS (ADAMS accession number ML091400198).

References:

RAI-25-1 Amendment 146, "Monticello Nuclear Generating Plant (MNGP) - Issuance of Amendment for the Conversion to the Improved Technical Specifications with Beyond-Scope Issues (TAC Nos. MC7505, MC7597 through MC761 1, And MC8887)" dated June 5, 2006.

L-MT-1 0-049 Page 6 of 7 NRC RAI 26 (Partial)

Please provide the following information relating to the Monticello MELLLA+ operation:

(1) details to obtain a final core loading pattern including procedure, guidance, criteria, and approved methodologies used for this analysis.

(2) when the final or reference core loadingpattern will be available for analyzing the cycle-specific operatinglimits listed in the Table of Section 2.2.

(3) when the final reload analysis report will be available for parameterslisted in Sections 2.3, 2.4, and 2.5 in the reload analysis report.

Response

The responses to this RAI as stated are contained in Attachment 2.

In addition to the questions above, the NRC staff (during a conference call on July 19, 2010) requested the core design for the Monticello Cycle 26 core. The Cycle 26 core design is provided on the following page as requested.

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