ML11300A226
| ML11300A226 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/25/2011 |
| From: | Franke J Progress Energy Co, Progress Energy Florida |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 3F1011-08, TAC ME6527 | |
| Download: ML11300A226 (10) | |
Text
Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 October 25, 2011 3F1011-08 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Crystal River Unit 3 - Feedwater Line Break Overpressure Protection Analysis to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 Extended Power Uprate LAR and LAR Approval Schedule (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. MLI 12070659)
- 2. CR-3 to NRC letter dated October 11, 2011, "Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)
- 3. Email from S. Lingam (NRC) to D. Westcott (CR-3) dated August 4, 2011, "CR-3 EPU LAR - Draft RAIs for Acceptance Review from SRXB (TAC No.
ME6527)"
Dear Sir:
By letter 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 (RTP) level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt. On August 4, 2011, via electronic mail, the NRC provided a request for additional information (RAI) related to Extended Power Uprate (EPU) accident and transient analyses needed to support the Reactor Systems Branch (SRXB) acceptance review of the CR-3 EPU License Amendment Request (LAR). By teleconferences on October 3, 2011 and October 6, 2011, CR-3 discussed the RAI with the NRC to confirm an understanding of the information being requested regarding the CR-3 overpressure protection analyses. On October 11, 2011, CR-3 provided a response to the SRXB Acceptance Review RAI which included a regulatory commitment to provide a feedwater line break (FWLB) overpressure protection analysis., "Feedwater Line Break Overpressure Protection Analysis to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 EPU LAR," provides an overview of the FWLB overpressure protection analysis and a comparison to the acceptance criteria of NUREG-0800, Section 5.2.2, "Overpressure Protection." To satisfy the regulatory commitment included in the SRXB Acceptance Review RAI response letter dated October 11, 2011 (Reference 2), an enclosure, "AREVA ANP-3052 - CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip, Revision 0," provides a summary of the CR-3 EPU FWLB overpressure protection analysis and includes key analysis input assumptions and Reactor Coolant System pressure results.
Additionally, CR-3 plans to resume power operation following completion of the CR-3 containment repair activities.
Due to the time required for the CR-3 containment repair, installation of the remaining EPU modifications is expected to occur during the current extended Progress Energy Florida, Inc.
Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428
U.S. Nuclear Regulatory Commission Page 2 of 3 3F101 1-08 outage period. As a result, FPC requests NRC review and approval of the proposed EPU license amendment to increase RTP to 3014 MWt within 24 months from the CR-3 EPU LAR submittal to support scheduling of the EPU implementation plan which is expected to occur during the plant startup following completion of the CR-3 containment repair activities.
FPC remains committed to providing sufficient and timely support of the NRC technical review of the CR-3 EPU LAR to aid in meeting the desired approval schedule. Attachment 2, "Additional Basis for the CR-3 EPU LAR Approval Schedule," provides additional basis for requesting the CR-3 EPU license amendment within 24 months from submittal.
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.
1S, J
A. ranke
- ice President CCrsI vrystal River Nuclear Plant JAF/gwe Attachments
- 1.
Feedwater Line Break Overpressure Protection Analysis to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 EPU LAR
- 2.
Additional Basis for the CR-3 EPU LAR Approval Schedule
Enclosure:
AREVA ANP-3052 - CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip, Revision 0 xc:
NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact
U.S. Nuclear Regulatory Commission Page 3 of 3 3F1011-08 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.
Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this J,5" day of jd,)(L*!*
, 2011, by Jon A. Franke.
Signature of Notary Public State of I
-CAROLYNE. PORTMANN Commission # DD 937553 Expires March 1. 2014 (Print, type, or stamp Commissioned Name of Notary Public)
Personally Produced Known V___
-OR-Identification
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 /LICENSE NUMBER DPR-72 ATTACHMENT 1 FEEDWATER LINE BREAK OVERPRESSURE PROTECTION ANALYSIS TO SUPPORT NRC REACTOR SYSTEMS BRANCH ACCEPTANCE REVIEW OF THE CR-3 EPU LAR
U. S. Nuclear Regulatory Commission 3F1011-08 Page 1 of 3 FEEDWATER LINE BREAK OVERPRESSURE PROTECTION ANALYSIS TO SUPPORT NRC REACTOR SYSTEMS BRANCH ACCEPTANCE REVIEW OF THE CR-3 EPU LAR On August 4, 2011, via electronic mail, the NRC provided a request for additional information (RAI) related to Extended Power Uprate (EPU) accident and transient analyses needed to support the Reactor Systems Branch (SRXB) acceptance review of the Crystal River Unit 3 (CR-3) EPU License Amendment Request (LAR). By teleconferences on October 3, 2011 and October 6, 2011, CR-3 discussed the RAI with the NRC to confirm an understanding of the information being requested regarding the CR-3 overpressure protection analyses. On October 11, 2011, CR 3 provided a response to the SRXB Acceptance Review RAI which included a regulatory commitment to provide a feedwater line break (FWLB) overpressure protection analysis (Reference 1).
This attachment provides an overview of the FWLB overpressure protection analysis and a comparison to the acceptance criteria of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," (SRP), Section 5.2.2, "Overpressure Protection."
The EPU overpressure protection analysis for the CR-3 most limiting postulated FWLB accident was performed to ensure the maximum primary system pressure remains within 120% of design pressure assuming a failure of the Reactor Coolant System (RCS) high pressure Reactor Protection System (RPS) trip.
The EPU analysis demonstrates compliance with the reactor coolant pressure boundary (RCPB) pressure acceptance criterion of 120% assuming a double-ended guillotine FWLB; a very low-probability event.
Limiting Primary System Overpressure Infrequent Event As described in Section 2.8.5.0, "Non-LOCA Analyses Introduction," of the CR-3 EPU Technical Report (Reference 2, Attachment 7), CR-3 does not consider any current licensing basis (CLB) accident or transient an infrequent event (i.e., American Nuclear Society (ANS)
Condition III). The CR-3 CLB FWLB accident is considered an ANS Condition IV accident; an event not expected to occur, but postulated because the consequences would include the potential for the release of significant amounts of radioactive material.
Nevertheless, a FWLB overpressure protection analysis was performed at EPU conditions for the most limiting postulated FWLB accident assuming a double-ended guillotine break and a failure of the RCS high pressure RPS trip at the request of the NRC Staff. This overpressure protection analysis confirms that the total relieving capacity of the pressurizer safety valves (PSVs) is sufficient to limit the maximum RCS pressure to less than 120% of design pressure.
The peak primary system pressure criterion of 120% of design pressure is consistent with the NRC Staff feedback received during the October 6, 2011 teleconference; SRP 5.2.2 Criterion 3.B related to the most severe infrequent event appears to be inconsistent with the preponderance of SRP 5.2.2 and should be 120% of design pressure.
To satisfy the regulatory commitment included in the SRXB Acceptance Review RAI response letter dated October 11, 2011 (Reference 1), an enclosure, "AREVA ANP-3052 - CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip, Revision 0," provides a summary of the CR-3 EPU FWLB overpressure protection analysis and includes key analysis input assumptions and RCS pressure results.
U. S. Nuclear Regulatory Commission 3F1011-08 Page 2 of 3 SRP Section 5.2.2 Comparison The CR-3 EPU overpressure protection analysis for the FWLB accident was performed considering the acceptance criteria of SRP Section 5.2.2 with the following exceptions:
The analysis does not include uncertainty limits on all system and core parameters as specified in SRP Section 5.2.2, Criterion 3.B.ii. Nominal values are assumed for some initial system parameters as allowed by an NRC approved methodology; BAW-10193NP-A "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water Reactors," (Reference 3). For example, a nominal value is assumed for initial RCS pressure in the enclosed FWLB overpressure protection analysis. A conservative assumption for the RCS high pressure Diverse Scram System (DSS) trip is used and includes instrument uficertainty plus additional engineering margin of approximately 82% of the instrument uncertainty.
This additional engineering margin, coupled with maximum lift tolerance conservatively added to the relief valve lift settings and other conservative boundary conditions (e.g., reactivity coefficients and inserted rod worth), obviates the need for initial RCS pressure to be adjusted for instrument uncertainty. The NRC approved the use of RELAP5/MOD2-B&W safety analyses code for non-Loss of Coolant Accident (LOCA) events as stated in the associated Safety Evaluation dated October 15, 1999 (Reference 4) and, as allowed by 10 CFR 50.34(h)(2), is considered an acceptable method of complying with applicable regulatory requirements.
" CR-3 does not have a second safety-grade signal from the RPS to initiate a reactor trip in the RCS overpressure protection analyses as specified in SRP Section 5.2.2, Criterion 3.B.iii.
CR-3 overpressure protection analyses do not postulate a failure of an RPS trip on RCS high pressure and is beyond CR-3 design and licensing basis. However, since the initial design and licensing of CR-3, several related plant improvements have been made for defense-in-depth. In response to implementation of the Anticipated Transient Without Scram (ATWS) rule requirements, 10 CFR 50.62(c)(2), CR-3 added a DSS. The DSS is independent of RPS and serves as a backup to the RPS RCS high pressure trip ensuring a reactor trip via a diverse method on overpressure ATWS transients. The DSS trips only regulating rods (i.e., Rod Groups 5-7) and is not safety-related. Although the DSS is not designed to maintain the peak pressure below 110% of the RCPB design pressure, the NRC concluded in 1989 (Reference 5) that the DSS design was in compliance with the ATWS rule. The DSS does not satisfy the SRP as a means of complying with the acceptance criterion of a second safety-grade RPS signal. However, the postulated failure of the RPS is fundamentally equivalent to an ATWS event. Thus, as allowed by 10 CFR 50.34(h)(2), the DSS constitutes an acceptable method of complying with the applicable regulatory requirements of providing backup RCS overpressure protection. Analysis results of the FWLB accident, assuming a failure of the RPS RCS high pressure trip, indicate a peak primary system pressure greater than 110% of the RCPB design pressure but less than 120% of the RCPB design pressure which is less than the value for unexpected system excess pressure transients specified in the American Society of Mechanical Engineers (ASME) Code Article NB-7000; calculated value required to exceed ASME Service Level C stress limits.
U. S. Nuclear Regulatory Commission 3F1011-08 Page 3 of 3 References
- 1.
CR-3 to NRC letter dated October 11, 2011, "Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Reactor Systems Branch Acceptance Review of the CR 3 Extended Power Uprate LAR" (TAC No. ME6527).
- 2.
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).
- 3.
BAW-10193NP-A "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water Reactors," Framatome Technologies Group, Lynchburg, Virginia, January 2000.
- 4.
NRC to Framatome Technologies letter dated October 15, 1999, Safety Evaluation of Topical Report BAW-10193P, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed PWRs," (TAC No. M93346).
- 5.
NRC to CR-3 letter dated April 19, 1989, "Crystal River Unit 3 - Review of Final ATWS Design," (TAC No. 59085).
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 /LICENSE NUMBER DPR-72 ENCLOSURE AREVA ANP-3052 - CR-3 EPU FEEDWATER LINE BREAK ANALYSIS WITH FAILURE OF FIRST SAFETY GRADE TRIP, REVISION 0
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ATTACHMENT 2 ADDITIONAL BASIS FOR THE CR-3 EPU LAR APPROVAL SCHEDULE
U. S. Nuclear Regulatory Commission 3F1011-08 Page 1 of 1 ADDITIONAL BASIS FOR THE CR 3 EPU LAR APPROVAL SCHEDULE By letter 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 (RTP) level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt. FPC requests NRC review and approval of the proposed Extended Power Uprate (EPU) license amendment to increase RTP to 3014 MWt by June 2013 with implementation occurring during the plant startup following completion of CR-3 containment repair activities.
The requested EPU license amendment approval date is crucial to CR-3 commercial planning and scheduling and supports an integrated startup test plan coincident with post-maintenance and startup testing required following containment repair activities.
As containment repair and plant startup testing schedules are finalized, FPC remains committed to frequent communication with the NRC Staff regarding the status of containment repair in order to project a more precise approval date for the proposed EPU license amendment.
In October 2009, during detensioning of the reactor building tendons in order to replace the once-through steam generators, FPC discovered delamination of the reactor building. Following initial repairs to the reactor building, subsequent containment structure delamination occurred during tendon retensioning in March 2011. As a result, an extended outage period was required to effect repairs to the containment structure before resuming commercial operation.
CR-3 plans to install the remaining EPU modifications during the current extended outage period. Implementing EPU during the plant startup following completion of containment repair activities will require extensive planning and scheduling to support an integrated startup test plan and required post-maintenance testing activities, including the EPU modifications. As a result, it is important that a technical review of the CR-3 EPU License Amendment Request (LAR),
including safety-related plant modifications assumed in EPU analyses, be completed and issues resolved to support installation of these modifications with minimum variances.
Installing the remaining EPU modifications during the current extended outage period provides several safety benefits, including but not limited to:
a new Inadequate Core Cooling Mitigation System will replace the current manual actions required upon a loss of subcooling margin with automatic actions, thereby reducing the reliance on time critical manual operator actions for event mitigation; the new Low Pressure Injection (LPI) System hot leg injection line will provide a more reliable method of core mixing and boron concentration control during core uncovery events thereby eliminating the single active failure exemption associated with existing boron precipitation mitigation methods; the addition of a LPI System cross-tie line will improve mitigation of certain spectra of Loss of Coolant Accidents (LOCAs); and High Pressure Injection System modifications and the addition of a Fast Cooldown System will improve small break LOCA mitigation.
In addition, implementing EPU upon resuming commercial operation will obviate the need to perform a separate mid-cycle shutdown and subsequent startup to implement EPU at CR-3, thereby avoiding an unnecessary plant transient.
Based on these considerations, FPC requests NRC review and approval of the proposed EPU license amendment to increase RTP to 3014 MWt within 24 months from the CR-3 EPU LAR submittal.