3F0912-03, Response to Second Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (Eivb) Technical Review of the CR-3 Extended Power Uprate LAR

Response to Second Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (Eivb) Technical Review of the CR-3 Extended Power Uprate LAR

ML12272A344
Person / Time
Site:	Crystal River
Issue date:	09/27/2012
From:	Franke J Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0912-03, TAC ME6527
Download: ML12272A344 (9)

Text

PDuke WEnergy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 September 27, 2012 3F0912-03 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Second Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (EIVB) 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" (ADAMS Accession No. ML112070659)

2. Email from S. Lingam (NRC) to P. Rose and D. Herrin (CR-3) dated August 2, 2012, "Crystal River, Unit 3 EPU LAR - Draft RAIs from EVIB (TAC No.

ME6527)"

3. NRC to CR-3 letter dated August 16, 2012, "Crystal River Unit 3 Nuclear Generating Plant - Request For Additional Information For Extended Power Uprate License Amendment Request (TAC No. ME6527)" (ADAMS Accession No. ML12219A191)

Dear Sir:

By letter dated June 15, 2011, Florida Power Corporation (FPC) 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 August 2, 2012, via electronic mail, the NRC provided a draft request for additional information (RAI) needed to support the EIVB technical review of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR) (Reference 2). By teleconference on August 9, 2012, FPC discussed the draft RAI with the NRC to confirm an understanding of the information being requested. On August 16, 2012, the NRC provided a formal RAI required to complete its evaluation of the CR-3 EPU LAR (Reference 3).

The attachment, "Response to Second Request for Additional Information - Vessels and Internals Integrity Branch Technical Review of the CR-3 EPU LAR," provides the CR-3 formal response to the RAI.

This correspondence contains no new regulatory commitments.

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

U.S. Nuclear Regulatory Commission 3F0912-03 Page 2 of 3 If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.

Jf A.-ranke Vice President Crystal River Nuclear Plant JAF/krw

Attachment:

Response to Second Request for Additional Information - Vessels and Internals Integrity Branch Technical Review of the CR-3 EPU LAR xc:

NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact

U.S. Nuclear Regulatory Commission 3F0912-03 Page 3 of 3 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation; 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.

J A. Franke ice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this a___7

_,2012, by Jon A. Franke.

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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT RESPONSE TO SECOND REQUEST FOR ADDITIONAL INFORMATION - VESSELS AND INTERNALS INTEGRITY BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR

U.S. Nuclear Regulatory Commission Attachment 3F0912-03 Page 1 of 5 RESPONSE TO SECOND REQUEST FOR ADDITIONAL INFORMATION - VESSELS AND INTERNALS INTEGRITY BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR By letter dated June 15, 2011, Florida Power Corporation (FPC) 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 August 2, 2012, via electronic mail, the NRC provided a draft request for additional information (RAI) needed to support the Vessels and Internals Integrity Branch (EIVB) technical review of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR). By teleconference on August 9, 2012, FPC discussed the draft RAI with the NRC to confirm an understanding of the information being requested. On August 16, 2012, the NRC provided a formal RAI required to complete its evaluation of the CR-3 EPU LAR. For tracking purposes, each item related to this RAI is uniquely identified as EIVB X-Y, with X indicating the RAI set and Y indicating the sequential item number.

EVIB REQUEST FOR ADDITIONAL INFORMATION EVIB-1 (EIVB 2-1)

Pressure-Temperature (P-T) Limit Curves:

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix G,Section IV.A, "Pressure-Temperature Limits and Minimum Temperature Requirements," indicates that the pressure-retaining components of the reactor coolant pressure boundary (RCPB) that are made of ferritic materials must meet the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section III, supplemented by the additional requirements set forth in 10 CFR 50, Appendix G,Section IV.A.2. Therefore, 10 CFR 50, Appendix G requires that P-T limits be developed for the ferritic materials in the reactor vessel (RV) beltline (neutron fluence > 1 x 1017 neutrons/square centimeter (n/cm2), Energy (E) > 1 mega electron volt (MeV)), as well as ferritic materials not in the RV beltline (neutron fluence

< 1 x 1017 n/cm 2, E > 1 MeV). Further, 10 CFR Part 50, Appendix G requires that all RCPB components must meet the ASME Code,Section III requirements. The relevant ASME Code requirement that will affect the P-T limits is the lowest service temperature requirement for all RCPB components specified in Section III, Paragraph NB-2332(b).

The P-T limit calculations for ferritic RCPB components that are not RV beltline shell materials may define P-T curves that are more limiting than those calculated for the RV beltline shell materials due to the following factors:

1.

RV nozzles, penetrations, and other discontinuities have complex geometries that may exhibit significantly higher stresses than those for the RV beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the reference temperature (RTNDT) for these components is not as high as that of RV beltline shell materials that have simpler geometries.

2.

Ferritic RCPB components that are not part of the RV may have initial RTNDT values, which may define a more restrictive lowest operating temperature in the P-T limits than those for the RV beltline shell materials.

U.S. Nuclear Regulatory Commission Attachment 3F0912-03 Page 2 of 5 Based on the above, please describe how the current P-T limit curves at 27.5 effective full power years for CR-3, and the methodology used to develop these curves, considered all RV materials (beltline and non-beltline) and the lowest service temperature of all ferritic RCPB materials, consistent with the requirements of 10 CFR Part 50, Appendix G in the EPU application.

Response

As required by CR-3 Improved Technical Specification (ITS) 5.6.2.19, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," the P-T limits were developed in accordance with the requirements of 10 CFR 50 Appendix G utilizing the analytical methods specified in Babcock and Wilcox (B&W) Topical Report BAW-10046A, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G," (Reference 2). The NRC safety evaluation report associated with BAW-10046A, Revision 2, states: "BAW 10046, Rev. 2 describes acceptable methods for the development of allowable pressure-temperature limits for normal operation and for test conditions to assure the prevention of non-ductile fracture. It may be referenced in future applications..." (Reference 3).

As noted in the Results of Section 2.1.2, "Pressure and Temperature Limits and Upper-Shelf Energy," of the CR-3 EPU Technical Report (Reference 1, Attachments 5 and 7):

"Using the post-EPU reactor vessel fluence projections, the current P/T limit curves are bounding for CR-3 up to 27.5 EFPY."

This statement means that the applicability of the current P-T limits was adjusted from 32 EFPY to 27.5 EFPY based on changes in reactor pressure vessel (RPV) fast neutron fluence (neutron E > 1.0 MeV) as a result of operation at EPU conditions. As of the end of CR-3 operating Cycle 16, the currently calculated EFPYs is 22.8.

The methodology specified in BAW-10046A, Revision 2, includes the ferritic components in reactor coolant pressure boundary (RCPB) locations other than the RPV beltline region in the development of the P-T limits.

Section 5.1, "Composite Limit Curves," of BAW-10046A describes the process of determining the composite P-T limits (e.g., Figures 5-2 and 5-4) and includes the P-T limiting regions of the RCPB. The P-T limits for the RPV closure head, outlet nozzle, and beltline region were determined to be limiting.

The RCPB material evaluations supporting the P-T limits are based on the adjusted reference temperature (ART) values, which utilize cumulative fast neutron fluence values; however, the limiting RCPB locations are not necessarily limiting solely based on these cumulative fluence values. The evaluation performed for EPU conditions adjusted the bounding EFPY limit at which the current P-T limits, based on limiting ART values, remain valid. To determine the applicable EFPY associated with the end of the CR-3 RPV service life, peak E > 1 MeV neutron fluence rates were determined for welds and places of interest on the reactor vessel surface, with projected EFPYs based on power operation at the respective power rating, including EPU power level. Projected cumulative RPV fluence values are determined based on: 1) current cumulative fluence values, 2) fluence rates based on expected core designs, and 3) time to reach the target (e.g., a specific EFPY value or RPV end of service life).

The current CR-3 P-T limits are based on the limiting RPV beltline welds which have ART values of 213'F at the 1/4T wall depth and approximately 144°F at the 3/4T wall depth. A calculation was performed considering EPU conditions to determine the minimum EFPY at

U.S. Nuclear Regulatory Commission Attachment 3F0912-03 Page 3 of 5 which the ART values, supporting the current P-T limits curves, remain valid. To ensure a different RPV beltline location did not become limiting, all beltline materials were considered.

This ART calculation was based on best estimate RCS chemistry conditions, surveillance material credibility assessments, and the most recent fluence information. It was determined that the ART calculation supporting the current P-T limits remains valid until at least 27.5 EFPY.

While the focus of the RCPB material evaluations for EPU operation was associated with fluence at the RPV beltline locations, the RPV fluence at the lower nozzle belt forging to the outlet nozzle forging weld and the lower RPV shell to the Dutchman circumferential weld were also determined.

The projected fluence for these welds is currently below the threshold for irradiation shift consideration, 1.0 x 1017 n/cm 2; and, considering EPU conditions, continues to be less than this threshold fluence through approximately 30 EFPY and 36 EFPY, respectively.

Therefore, irradiation shift is not applicable to the non-beltline ferritic RPV materials and the generic ART values for the RPV nozzle corner region and the RPV closure head reported in BAW-10046A remain valid during the remaining period in which the existing P-T limits have been determined acceptable; approximately 4.5 EFPY.

Also, as indicated in BAW-1 0046A, the lowest service temperature for the RCS piping is 150'F, based on RTNDT + 100°F, and the lowest service temperature for the control rod drive mechanisms (CRDMs) is 1 00'F. The effects of the additional fluence, as a result of operation at EPU conditions on RCS piping and CRDMs, does not result in these components becoming limiting since the lowest service temperature of the RPV head is greater than the lowest service temperature for the RCS piping and CRDMs; the lowest service temperature of the RPV head is 180'F based on an RTNDT of 60'F, where the lowest service temperature of the RPV head is represented by RTNDT + 120 0F.

As implied by 10 CFR 50, Appendix H,Section III.A, vessel embrittlement due to irradiation damage is not expected to occur until a fluence accumulation of 1.0 x 1017 n/cm 2 (E > 1.0 MeV) is reached. Therefore, RCPB material evaluations have not historically considered embrittlement from irradiation damage in the ferritic components adjacent to the RPV beltline (i.e., non-beltline region components) until a fluence accumulation of 1.0 x 1017 n/cm 2 (E > 1.0 MeV) is reached.

The industry has begun to address the embrittlement of areas of geometric discontinuity in the RPV, including inlet and outlet nozzles and transition regions. It is anticipated that the industry will develop additional methodology for this recent issue of concern for components with embrittlement for components with fluence less than 1.0 x 1017 n/cm 2 (E > 1.0 MeV). However, as described herein, the methodology specified in BAW-10046A, Revision 2, includes the ferritic components in RCPB locations other than the RPV beltline region in the development of the P-T limits and provides composite P-T limit curves based on the limiting regions.

The effects of the additional fluence, as a result of operation at EPU conditions, on RCPB components have been evaluated and it has been determined that the current P-T limits remain acceptable for EPU operation up to at least 27.5 EFPY.

EIVB-2 (EIVB 2-2)

RV Internals:

CR-3 participated in the industry effort for providing inspection and evaluation guidelines for plants to ensure integrity of RV internals. The product of this industry effort is the Electrical Power Research Institute's technical report, "Materials Reliability Program (MRP-227-A):

Pressurized Water Reactor [PWR] Internals Inspection and Evaluation Guidelines," December

U.S. Nuclear Regulatory Commission Attachment 3F0912-03 Page 4 of 5 2011 (ADAMS Accession No. ML120170453). Sections 7.2 and 7.3 of the MRP 227-A have specified requirements related to RV internals to be executed during the current 40-year license:

(1) MRP-227-A Section 7.2, "Aging Management Program Requirement,"

requires, "Each commercial U.S. PWR unit shall develop and document a program for management of aging of reactor internal components within thirty-six months following issuance of MRP-227-Rev. 0 (that is, no later than December 31, 2011)."

Please confirm that you have met this MRP-227-Rev. 0 requirement by the date specified in the parentheses to support the CR-3 EPU application.

(2) MRP-227-A Section 7.3, "Reactor Internals Guidelines Implementation Requirement," requires, "Implementation of these guidelines [MRP-227-A, Tables 4-1 through 4-9 and Tables 5-1 through 5-3] is to take effect 24 months following issuance of MRP-227-A (that is, no later than December 31, 2013).

Implementation means performance of inspections of applicable components within the time frame specified in the guidance provided in the applicable tables."

Please confirm that you will meet the above MRP-227-A requirement by the date specified in the parentheses to support the CR-3 EPU application.

Response

WCAP-17113-NP, "PWR Vessels Internal Program Plan for Aging Management for Reactor Internals at Crystal River Unit 3," was developed for CR-3 to manage the effects of aging on reactor vessel internals. WCAP-17113-NP was issued in September 2009, which was within the "mandatory" thirty-six month following issuance of Revision 0 of MRP-227 (i.e., prior to December 31, 2011).

Most of the applicable MRP-227 inspection requirements identified in the tables associated with the B&W plants, Tables 4-1, 4-4, 4-7, and 5-1, are not due until after extension of the CR-3 operating license. In addition, FPC has already performed the applicable MRP-227 inspections that would have been due prior to extension of the CR-3 operating license with one deviation.

FPC has deferred the ultrasonic testing inspection of the accessible RPV lower core barrel bolts until the next removal of the core support structure, which cannot occur prior to December 31, 2013 due to the CR-3 containment repair activities. Therefore, with the exception of inspecting the lower core barrel bolts and their locking devices, FPC has performed the applicable inspections in accordance with the guidelines provided in the MRP-277; within the "needed" twenty-four months following issuance of MRP-227-A (i.e., prior to December 31, 2013).

References

1.

FPC to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate."

(ADAMS Accession No. ML112070659)

2.

B&W Topical Report BAW-10046A, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G," Revision 2, dated June 1986.

U.S. Nuclear Regulatory Commission Attachment 3F0912-03 Page 5 of 5

3.

NRC to B&W letter dated April 30, 1986, "Acceptance for Referencing of Licensing Topical Report BAW-10046, Rev. 2 B&W Owners Group Materials Committee 'Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G."' (ADAMS Accession No. 8607230131)