3F0412-04, Response to Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch (Emcb) Technical Review of the CR-3 Extended Power Uprate LAR

Response to Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch (Emcb) Technical Review of the CR-3 Extended Power Uprate LAR

ML12097A246
Person / Time
Site:	Crystal River
Issue date:	04/04/2012
From:	Franke J Progress Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0412-04, TAC ME6527
Download: ML12097A246 (6)

Text

Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 April 4, 2012 3F0412-04 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch (EMCB) Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)

References:

1. CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate" (Accession No. ML112070659)

2. NRC to CR-3 letter dated March 2, 2012, "Crystal River Unit 3 Nuclear Generating Plant - Request For Additional Information For Extended Power Uprate License Amendment Request (TAC No. ME6527)" (Accession No. ML12052A130)

Dear Sir:

By letter dated June 15, 2011, Florida Power Corporation, doing business as Progress Energy Florida, Inc., requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt (Reference 1). On March 2, 2012, the NRC provided a request for additional information (RAI) required to support the EMCB technical review of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR) (Reference 2).

The attachment, "Response to Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch Technical Review of the CR-3 EPU LAR," provides the formal response to the RAI needed to support the EMCB technical review of the CR-3 EPU LAR.

This correspondence contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.

Sinmce el JJo Jo

.ranke ice President Crystal River Nuclear Plant JAF/gwe

Attachment:

Response to Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch Technical Review of the CR-3 EPU LAR xc:

NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact Progress Energy Florida, Inc.

j'Z Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428

U.S. Nuclear Regulatory Commission Page 2 of 2 3F0412-04 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief..

/

on A. Franke Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this day of 2012, by Jon A. Franke.

Signature of Notary Public State of Florida l,,t'*g".. CAROLYN E. PORTMANN Commission # DD 937553 4i Epires March 1,2014 (Print, type, or stamp Commissioned Name of Notary Public)

Personally Produced Known

//-OR-Identification

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 /LICENSE NUMBER DPR-72 ATTACHMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TO SUPPORT NRC MECHANICAL AND CIVIL ENGINEERING BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR

U. S. Nuclear Regulatory Commission Attachment 3F0412-04 Page 1 of 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TO SUPPORT NRC MECHANICAL AND CIVIL ENGINEERING BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR By letter (Reference 1) dated June 15, 2011, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt. On March 2, 2012, the NRC provided a request for additional information (RAI) required to support the Mechanical and Civil Engineering Branch (EMCB) technical review of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR). For tracking purposes, each item related to this RAI is uniquely identified as EMCB X-Y, with X indicating the RAI set and Y indicating the sequential item number.

EMCB RAIs

15.

(EMCB 1-1)

The control rod drive mechanism (CRDM) analysis discussion on page 2.2.2.4-1 of Attachment 5 of the original license amendment request (LAR) indicates that the extended power uprate (EPU) operating temperature of 608.7 'F slightly exceeds the design analysis temperature of 608

'F.

The discussion does not indicate that the licensee intends to revise the design report to reflect this change. Please provide assurances that the CRDM analysis either has been or will be revised to include a discussion of the new operating temperature. The revision needs to include a statement that Code requirements continue to be met under the new condition which will be present at the proposed EPU power level.

Response

A revision to the CRDM design report is not necessary since thermal changes to the CRDMs do not adversely impact the conditions or results of this report as a result of operation at EPU conditions.

The design temperature of the CR-3 CRDMs is 650'F and it is not altered or exceeded as a result of EPU conditions. The calculated CRDM operating temperature increases to 608.77F and is slightly higher than the CRDM operating temperature assumed in the design specification; 608'F.

A separate EPU evaluation of reactor coolant pressure boundary components assessed the impact of a 0.77F increase above the CRDM operating temperature value in the design specification. The results of this evaluation are reported in Section 2.2.2, "Pressure-Retaining Components and Component Supports," of the EPU Technical Report (TR)

(Reference 1, Attachments 5 and 7).

As noted in Section 2.2.2.4, "Control Rod Drive Mechanism and Supports," of the EPU TR, the temperature increase of 0.7°F above the currently qualified CRDM operating temperature value of 6087F results in a negligible change in CRDM thermal stress. Furthermore, Section 2.2.2.4 notes that there are no changes to the Nuclear Steam Supply System design transients that would increase stress or fatigue in the CRDMs. As such, the evaluation concludes that the ASME Code stress and fatigue requirements continue to be met during operation at EPU conditions and the CRDM design report does not require an update as a result of operation at EPU conditions.

U. S. Nuclear Regulatory Commission Attachment 3F0412-04 Page 2 of 3

16.

(EMCB 1-2)

The steam generator base support evaluation discussion on page 2.2.2.5-2 of Attachment 5 of the original LAR states that the base supports are designed to the 2000 Addenda of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, Subsection NF.

However, neither Table 4-2 of the Crystal River Unit 3 (CR-3) final safety analysis report (FSAR) nor CR-3 FSAR Section 4.6.2.3 includes these criteria. The licensee needs to provide a reference to an existing reconciliation between the 2000 Addenda criteria and the criteria listed in the FSAR, or provide reconciliation between the two criteria.

Response

The integral steam generator base supports are not specifically addressed in the CR-3 FSAR.

FSAR Table 4-2 addresses the pressure vessel portions (primary and secondary sides) of the replacement once through steam generators (ROTSGs).

FSAR Section 4.2.6 addresses the foundations and lateral restraints for the steam generators.

As noted in American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, Sub-article IWA-4226, "Reconciliation of Design Requirements:"

When an item is designed to all requirements of a later Edition or Addenda of the Construction Code, or Section III when the Construction Code was not Section III, reconciliation beyond the design-related issues defined in IWA-4223, IWA-4224 and IWA-4225 is not required ASME Section XI, Sub-article IWA-4223, "Reconciliation of Components," requires evaluation of changes in weight, configuration, or pressure temperature ratings.

This evaluation was performed prior to installation of the CR-3 ROTSGs.

The ROTSG supplier determined that ASME Section XI, Sub-article IWA-4225, "Reconciliation of Parts, Appurtenances and Piping Subassemblies," was not applicable to the CR-3 ROTSGs for code reconciliation since parts, appurtenances, and piping sub-assemblies are fabricated to ASME BPV Code Edition 1998 and 2000 Addenda.

ASME Section XI, Sub-article IWA-4224, "Reconciliation of Materials," requires reconciliation of materials procured to later codes. This reconciliation was performed and documented in the CR-3 EPU ASME Design Report. The CR-3 EPU ASME Design Report summarizes the ASME Code (Reference 2) "design-by-analyses" of the ROTSGs considering EPU conditions and reconciles the analyses with the "As-Built" conditions of the ROTSGs, including reconciliation of the steam generator base supports. This report includes a Registered Professional Engineer certification that the design report is: "...complete and accurate and complies with the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Class 1, Subsections NB and NF, 1998 Edition, 2000 Addenda." Since the CR-3 ROTSG design meets the requirements of the later edition and addenda of the ASME BPV code, no further reconciliation is required.

U. S. Nuclear Regulatory Commission Attachment 3F0412-04 Page 3 of 3

17.

(EMCB 1-3)

The in-core instrumentation guide tubes evaluation discussion on page 2.2.6-2 of Attachment 5 of the original LAR indicates that no re-evaluation of thermal stresses is required based on the reactor T-cold decreasing by less than 1 'F. However, on page 2.2.2.3-1 of the LAR, the reactor T-hot temperature is listed as increasing by 6.6 'F with the EPU, giving an average reactor temperature increase of approximately 3 'F.

State the impact of the increases of T-hot and T-average on the structural integrity of the incore guide tubes and provide a justification for not re-evaluating the stresses in these tubes based on apparently more complex thermal conditions at the proposed EPU power level.

Response

As noted in the Introduction of Section 2.2.6, "Incore Instrumentation Guide Tubes," of the EPU TR (Reference 1, Attachments 5 and 7), the incore instrumentation nozzles are addressed in Section 2.2.2.3, "Reactor Vessel and Supports," and the non-pressure boundary portion of the incore instrumentation guide tubes internal to the reactor vessel are addressed in Section 2.2.3, "Reactor Pressure Vessel Internals and Core Supports."

The incore instrumentation guide tubes originate at the bottom head of the reactor vessel at the incore instrument nozzles and are routed to the incore closure assembly at the incore instrument tank. Refer to Figure 7-20, "Incore Monitoring Channel," of the CR-3 FSAR for a simplified structural diagram of the pressure retaining incore guide tube addressed in Section 2.2.6 of the EPU TR (Reference 1, Attachments 5 and 7). Due to this configuration, the pressure retaining incore guide tubes and incore closure assembly are only exposed to fluid at TCOLD conditions since the reactor coolant comes in contact with them prior to passing through the fuel assemblies and heating up to THOT. Based on this, an increase in THOT and TAVG has no thermal impact on the pressure retaining portion of the incore guide tubes. As such, the structural integrity of the pressure retaining incore guide tubes is not adversely affected as a result of operation at EPU conditions and additional thermal stress analyses associated with the incore guide tubes are not needed for EPU operation. Additionally, FPC evaluated the stress and fatigue of the reactor vessel internals considering the increase in THOT and TAVG. As indicated in Section 2.2.3 of the EPU TR (Reference 1, Attachments 5 and 7), the evaluation included the non-pressure boundary portion of the incore instrumentation guide tubes internal to the reactor vessel and concluded that stress and fatigue of the reactor vessel internals due to thermal loading are unaffected by operation at EPU conditions.

References

1.

CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate." (Accession No. ML112070659)

2.

ASME Boiler and Pressure Vessel Code, 1998 Edition, including 2000 Addenda.