ML14339A163
| ML14339A163 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 12/23/2014 |
| From: | Lisa Regner Plant Licensing Branch II |
| To: | Nazar M Nextera Energy |
| Saba F DORL/LPL2-2 301-415-1447 | |
| References | |
| TAC MF4490 | |
| Download: ML14339A163 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Mano Nazar President and Chief Nuclear Officer Nuclear Division NextEra Energy P.O. Box 14000 Juno Beach, FL 33408-0420 December 23, 2014
SUBJECT:
ST. LUCIE PLANT UNIT NO. 1 - RELIEF REQUEST NUMBER 8-ALTERNATIVE FROM PERFORMING VOLUMETRIC/SURFACE EXAMINATIONS OF REACTOR VESSEL CLOSURE HEAD COMPONENTS (TAC NO. MF4490)
Dear Mr. Nazar:
By letter dated July 24, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14206A939), as supplemented by letter dated September 26, 2014 (ADAMS Accession No. ML14273A011 ), Florida Power & Light Company (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI requirements at the St. Lucie Plant Unit No. 1 (SL-1).
Specifically, the licensee proposed to extend the frequency of the volumetric/surface examinations for the SL-1 reactor vessel closure head (RVCH) for approximately 3 years beyond the one inspection interval from the installation of the SL-1 replacement RVCH.
Pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Section 50.55a(a)(3)(i), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.
The NRC staff has reviewed Relief Request (RR) No. 8 and determined that the alternative method proposed by the licensee will provide an acceptable level of quality and safety for the examination frequency requirements of the reactor vessel closure head. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i). Therefore, the NRC staff authorizes the one-time use of RR No. 8 at SL-1 for the duration up to and including the 28th SL-1 refueling outage that is scheduled to commence in March 2018, which will occur in the fifth 1 0-year inservice inspection interval.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
If you have any questions, please contact Farideh E. Saba by phone at 301-415-1447, or by e-mail at Farideh.Saba@nrc.gov.
Docket No. 50-335
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via Listserv Lisa M. Regner, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO.8 -INSPECTION OF REACTOR VESSEL CLOSURE HEAD NOZZLES IN ACCORDANCE WITH ASME CODE CASE N-729-1 AS CONDITIONED BY 10 CFR 50.55a FLORIDA POWER AND LIGHT COMPANY ST. LUCIE PLANT UNIT NO. 1 DOCKET NO. 50-335
1.0 INTRODUCTION
By letter dated July 24, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14206A939), as supplemented by letter dated September 26, 2014 (ADAMS Accession No. ML14273A011), Florida Power & Light Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, associated with the examination frequency requirements of Code Case N-729-1 at St. Lucie Plant Unit No. 1 (SL-1).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR)
Section 50.55a(a)(3)(i), the licensee requested to use the proposed alternatives to the examination frequency of ASME Code Case N-729-1 on the basis that the alternative examination provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
The inservice inspection (lSI) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the U.S. Nuclear Regulatory Commission (NRC or the Commission).
Pursuant to 10 CFR 50.55a(g)(6)(ii), the Commission may require the licensee to follow an augmented lSI program for systems and components for which the Commission deems that added assurance of structural reliability is necessary. As required by the Code, in part, all licensees of pressurized water reactors shall augment their lSI with ASME Code Case N-729-1, "Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Enclosure Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," subject to conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section. Relief from 10 CFR 50.55a(g)(6)(ii)(D}, "Reactor vessel head inspections," is requested by the licensee.
As 10 CFR 50.55a(a)(3) states, alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the proposed alternative requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 Components Affected The affected components are ASME Class 1, Reactor Vessel Closure Head (RVCH)
Penetration Nozzle numbers 0-1 through 0-69, which are fabricated from lnconel SB-167 (Alloy 690) UNS N06690. The nozzle J-groove welds are fabricated from ERNiCrFe-7 (UNS N06052) and ENiCrFe-7 (UNS W86152}, 52/152 weld materials. The original SL-1 RVCH penetration nozzles, which were manufactured with Alloys 600/82/182 materials, were replaced with a new RVCH using Alloys 690/52/152 material for the penetration nozzles during the refueling outage that returned the plant to operation in December 2005.
3.2 lnservice Inspection Interval The licensee's current lSI interval is the fourth 1 0-year lSI interval that started in February 11, 2008, and continues through February 10, 2018. The proposed duration of the alternative will occur in the fifth 1 0-year lSI interval that is from February 11, 2018, to February 10, 2028.
3.3 ASME Code of Record The ASME Section XI Code of Record for the current fourth 1 0-year lSI interval at SL-1 that began February 11, 2008, and continues through February 10, 2018, is the 2001 Edition through the 2003 Addenda.
3.4 ASME Code and/or Regulatory Requirements Section 50.55a(g)(6)(ii)(D) of 10 CFR requires, in part, that licensees shall augment their lSI program in accordance with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(D). ASME Code Case N-729-1, Table 1, Inspection Item 84.40, requires volumetric/surface examination be performed within one inspection interval (nominally 10 calendar years) of its in service date for a replaced RVCH. The required volumetric/surface examinations need to be completed by December 2015 in order to fulfill the requirements of N-729-1.
3.5 Proposed Alternative The licensee proposes to delay the next required inspection for a period of approximately 3 years. The licensee proposes to accomplish the inspection in accordance with ASME Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D) during refueling outage 28 that is scheduled for March 2018. The NRC staff notes that the inspection date required by the regulations occurs in the plant's fourth lSI interval and that the proposed inspection will be accomplished during the plant's fifth lSI interval.
3.6 Licensee's Basis for Use of the Proposed Alternative The licensee's basis for use of the proposed alternative is based primarily on three topics of consideration. The first topic addresses the concept that the inspection interval in Code Case N-729-1 is based on primary water stress corrosion cracking (PWSCC) crack growth rates for Alloy 600/82/182. The second topic addresses a bare metal visual examination conducted in 2010, on the licensee's replacement RVCH, which had been replaced in 2005. The third topic addresses a plant-specific factor of improvement analysis conducted by the licensee.
In addressing its first basis for use of the proposed alternative, the licensee asserts that the inspection intervals contained in ASME Code Case N-729-1 for alloy 600/82/182 are based on reinspection years (RIY) equal to 2.25 and that this value is based on PWSCC crack growth rates as defined in the 751h percentile curve contained in MRP-55, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material," and MRP-115, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds." The licensee further asserts that the PWSCC crack growth rates of alloy 690/52/152 are significantly lower than those of alloy 600/82/182 and, therefore, merit a longer inspection interval. The licensee bases that assertion on a) the lack of cracking in other 690 components such as steam generators and pressurizers in the approximately 20 years that alloy 690 has been in service in these components; b) the failure to observe cracking in inspections already performed in replacement heads (9 of 40 replacement heads have been examined that include heads that operate at higher temperatures than the head under consideration); c) the similarity of the inspected heads to the head under consideration regarding configuration, manufactures, design, and operating conditions; and d) laboratory test data for alloys 690/52/152 as contained in MRP-375, "Technical Basis for Reexamination Interval for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles."
In addressing its second basis for use of the proposed alternative, the licensee stated that a bare metal visual examination was performed in 2010 on the SL-1 replacement RVCH in accordance with ASME Code Case N-729-1, Table 1, item B4.30. This visual examination was performed by Visual Testing (VT-2) qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. This examination did not reveal any indications of nozzle leakage (e.g., boric acid deposits) on the surface or near a nozzle penetration. The licensee also indicated that this examination will be performed again in the upcoming 26th refueling outage scheduled to commence in March 2015. Also, the licensee stated that no alternative examination processes are proposed to those required by ASME Code Case N-729-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). The VT-2 examinations and acceptance criteria as required by item ASME Code Case N-729-1, Table 1, item B4.30 are not affected by this request and will continue to be performed on a frequency not to exceed every 5 calendar years.
In addressing its third basis for use of the proposed alternative, the licensee made a plant specific calculation of the required factor of improvement in the crack growth rate of Alloy 690/52/152 as compared to the crack growth rate of alloy 600/82/182. In making this calculation, the licensee used the actual temperature of the head and conservatively assumed that calendar years were equal to effective full-power years. Based on this calculation, the licensee determined that a factor of improvement of 6.2 was required to meet the proposed and desired inspection interval of 13 calendar years. The licensee then proposed that because the required factor of improvement (6.2) was smaller than the factor of improvement of 20, which bounded most of the MRP-375 data for alloy 690/52/152, the use of a factor of improvement of 6.2 would not result in a reduction in safety and was, therefore, justified.
The licensee stated that its analysis showed significant margin to ensure that Alloy 690 nozzle base and Alloy 52/152 weld materials used in the SL-1 replacement RVCH provide for a reactor coolant system pressure boundary where the potential for PWSCC has been shown by analysis and by years of positive industry experience to be remote. As such, the licensee found the technical basis sufficient to ensure public health and safety by extending the inspection frequency of the RVCH nozzle at SL-1 from a maximum of 10 years to a new maximum of 13 years.
- 3. 7
NRC Staff Evaluation
In evaluating the technical sufficiency of the licensee's proposed alternative (i.e., a one-time extension of the volumetric/surface examination interval contained in ASME Code Case N-729-1 from 10 years to not longer than 13 years), the NRC staff considered each of the three aspects of the licensee's basis for use of the proposed alternative. The NRC staff found that the technical basis included by the licensee provided sufficient information for the NRC staff to review the proposed alternative.
Due to concerns about PWSCC, many PWR plants in the United States and overseas have replaced reactor vessel closure heads containing Alloy 600/182/82 nozzles with heads containing Alloy 690/152/52 nozzles. The inspection frequencies developed in Code Case N-729-1 for RVCH penetration nozzles using Alloy 600/182/82 were developed based, in part, on those material's crack growth rate equations documented in MRP-55 and MRP-115. The licensee's primary technical basis is to present crack growth rate data for the new more crack-resistant materials, Alloy 690/152/52, and demonstrate a factor of improvement of these materials versus the older Alloy 600/82/182 materials. This factor of improvement would then provide the basis for the extension of the lSI frequency requested by the licensee in its proposed alternative.
In evaluating the licensee's first technical basis for use of the proposed alternative, the NRC staff notes that the licensee uses MRP-375. This document, in part, summarizes numerous Alloy 690/152/52 crack growth rate data from various sources to develop factors of improvement for the crack growth rate equations provided in MRP-55 and MRP-115. While the NRC staff finds the licensee's assertions and/or interpretations to be reasonable, MRP-375 is not an NRC-approved document. As the NRC staff does not have sufficient time or resources to validate all of the data used by this document, the NRC staff does not consider it appropriate to use all of the data from this document to review the licensee's relief request. A more detailed review of the data provided in MRP-375 will be performed by an international group of experts as part of an Alloy 690 Expert Panel, which is currently scheduled to complete its review in the 2016-2017 timeframe. In the interim, the NRC staff review relied upon Alloy 690/152/52 crack growth rate data from two NRC contractors: Pacific Northwest National Laboratory (PNNL) and Argonne National Laboratory (ANL). This data is documented in a data summary report and can be found under ADAMS Accession No. ML14322A587. The NRC confirmatory research generally supports the contention that the crack growth rate of alloy 690/52/152 is more crack resistant but differs from the MRP-375 data in some respects.
The PNNL and ANL data summary report includes crack growth rate data up to approximately 20 percent cold work based on the observation of local strains in welds and weld dilution zone data. However, the NRC staff did not consider the weld dilution zone data in its assessment.
This is because the limited weld dilution zone data that is currently available has shown higher crack growth rates than are commonly observed for alloy 690/152/52 material. The high crack growth rates in weld dilution zones may be due to the reduced chromium present in these areas. The NRC staff chose to exclude the weld dilution zone data from this analysis due to the limited number of data points available, the variability in results, and the limited area of continuous weld dilution for flaws to grow through. For example, in the case of the highest measured crack growth rates, a flaw would have to travel in the heat affected zone of a J-groove weld along the low alloy steel head interface. It is not fully apparent to the NRC staff how accelerated crack growth in very small areas of weld dilution zone would result in a significantly increased probability of leakage or component failure during a relatively short extension of the required inspection interval. Exclusion of these data may be reevaluated as additional data become available, a better understanding of the existing data is obtained, or if a longer extension of the inspection interval is requested. Therefore, the NRC staff finds that the impact of these weld dilution zone crack growth rates on the change in volumetric inspection frequency, as requested by the licensee's proposed alternative, is not considered to be relevant for this specific relief request.
In evaluating the licensee's second basis for use of the proposed alternative, the NRC staff finds that the past bare metal visual examination on the head under consideration is a reasonable means to demonstrate the absence of leakage through the nozzle/J-groove weld prior to the time the examination was conducted. The NRC staff also finds that performance of future bare metal visual examinations in accordance with the code case is adequate to demonstrate the absence of leakage at or prior to the time the examinations are conducted. Finally, the NRC staff finds that the proposed alternative's frequency for bare metal visual examinations in conjunction with the new frequency of volumetric examinations is sufficient to provide reasonable assurance of the structural integrity of the RVCH.
In evaluating the licensee's third basis for use of the proposed alternative, the NRC staff found that the licensee's calculated factor of improvement of 6.2, to support an extension of the ASME Code Case N-729-1 inspection frequency of 2.25 RIY to 13 calendar years was acceptable by NRC staff calculation. The NRC staff also found that the application of a factor of improvement of 6.2 to the 751h percentile curves in MRP-55 and MRP-115 bounded essentially all of the NRC data included in the PNNL and ANL data summary report. Therefore, the NRC staff found that this analysis supports the concept that a volumetric inspection interval for the RVCH of not more than 13 calendar years does not pose a higher risk than that associated with an alloy 600/182/82 RVCH inspected at intervals of 2.25 RIY. Hence, the NRC staff found the licensee's technical basis to be acceptable.
Therefore, the NRC finds that the proposed alternative provides an acceptable level of quality and safety as required by 10 CFR 50.55a(a)(3)(i).
4.0 CONCLUSION
As set forth above, the NRC staff has determined that the alternative method proposed by the licensee in RR No. 8 will provide an acceptable level of quality and safety for the examination frequency requirements of the RVCH. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i).
Therefore, the NRC staff authorizes the one-time use of RR No. 8 at SL-1 for the duration up to, and including, the 28th SL-1 refueling outage that is scheduled to commence in March 2018, which will occur in the fifth 1 0-year lSI interval.
All other requirements of the ASME Code,Section XI, and 10 CFR 50.55a(g)(6)(ii)(D) for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: Margaret T. Audrain Date: December 23, 2014
- via e-mail OFFICE LPL 1-1/PM LPL12-2/PM LPL2-2/LAiT LPL2-2/LA DE/EPNB/BC* LPL2-2/BC(A)
NAME AChereskin FSaba LRonewicz BCiayton DAiley LRegner (FSaba for)
DATE 12/10/14 12/10/14 12/09/14 12/09/14 11/21/14 12/23/14