RNP-RA/13-0073, Response to NRC Request for Additional Information Related to the Pressurized Water Reactor Internals Program Plan for Aging Management of Reactor Internals
| ML13219A252 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/25/2013 |
| From: | Wheeler-Peavyhouse S Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/13-0073, TAC ME9633 | |
| Download: ML13219A252 (11) | |
Text
LE(DitK E Elacft M"an " 2 fbfftai.* SC 29M6 F: 8305N7 1319 10 CFR 54.21 Serial: RNP-RA/13-0073 JUL 2 52013 ATTN: Document Control Desk United States Nuclear Regulatory Commission Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE PRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FOR AGING MANAGEMENT OF REACTOR INTERNALS (TAC NO. ME9633)
Ladies and Gentlemen:
By letter dated May 29,2013, the NRC requested that Duke Energy Progress, Inc., formerly known as Carolina Power and Light Company, respond to a second request for additional Information (RAI) regarding the Aging Management Program for the Reactor Vessel Internals at Robinson Nuclear Power Plant. Duke Energy Progress's response to this RAI Is provided in the enclosure to this letter.
There are no regulatory commitments made In t*fs submittal. If you have any questions reading ths submittal, please contact Mr. R. Hightower at (843) 857-1329.
I declare under penalty of perjury that the foregoing ts true and correct.
Executed On: -, 5 Z1 Sincerely, Q.,
sharo A.WheelerPeavyhouse Manager - Support Services, - Nuclear A Q37
United States Nuclear Regulatory Commission Serial: RNP-RA/13-0073 Page 2 of 2 SAWP/am
Enclosure:
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE PRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FOR AGING MANAGEMENT OF REACTOR INTERNALS cc: Mr. V. M. McCree, NRC, Region 11 Ms. A. T. Siocholon, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP Unit No. 2
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/13-0073 Page 1 of 9 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE PRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FOR AGING MANAGEMENT OF REACTOR INTERNALS (TAC NO. ME9633) DOCKET NO. 50-261 NCi REQOU0Ego FORADIT!NL INFORMATION (RMI By letter to the U.S. Nuclear Regulatory Commission (NRC) dated September 26, 2012, (Agencywide Documents Aces and Management System (ADAMS) Accession No. ML12278A398 and ML12278A390), Duke Energy Progress, Inc., formerly known as Carolina Power and Light Company, submitted an aging management program (AMP) for the reactor vessel intermals for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No.2. Based on the review of HBRSEP's AMP conducted thus far, the NRC staff developed a first RAI that was provided to the licensee on March 27, 2013. The licnsee provided its response to the first RAI on May 23, 2013. This enclosure contains HBRSEP's responses to the aforementioned second RAt.
RAI 2-1: RVI Componenft RAI 2-1:
a) Identify all RVI components in the HIBRSEP plant design that are defined in the current lansing basis (CLG)as ASME Section XI, Examination Category B-N-3 core support struure components. For these components, identify which of the four inspection categories in MRP-227-A is applicable to the component.
b) If a component identified in the reply to Part (a) Is defined as either a "Primary Category" or "Expansion Category" component, identify any differences between the inspections that would be performed on the components under MRP.227-A versus the plant's ASME Section Xl inservice Inspection program.
c) For the components identified In Part (b), clarify how the differences in inspection bases for these components will be reconciled consistent with the CL8 for the facility.
RpoNu W- to We 2-1:
a) Robinson has submitted the fourth interval in-service Inspection (ISI) program [1] to the NRC in [83, as required by 10 CFR 50.55(a). The fourth interval has ended; therefore Robinson is currently in the fifth interval. The components listed in Table 1 are identified In the updated, fifth interval, program [23 as ASME Section XI, Examination Category B-N-3 components. The MRP-227-A inspection category [(9 is indicated.
b) The differences between the inspecions that would be perfomed on the components under MRP-227-A versus the plant ASME Section Xl Inservice inspection program are summatized in Table 2.
c) For the components identified In Part (b), the inspections that are specified in MRP-227-A are either cons t with or more rigorous than what is specified as part of the ASME Section Xl inspections.
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA13-0073 Page 2 of 9 TABLE 1 KMA obkWon ASWE Sc901 V, C w 0x8."u -N-.3 Components Componbt MRP-227-A Inspection u ,rcore pe algn"met keyways 0, 90,,1w0, ?70° No 4fm M u upper core plate bottom surfaces Expansion UPPer core Plate aligmnt pins 0°, So°, 180°, 270° Existing top former plate Primary lower core pl:te Existin lower core piLte fuel apfnmewtpins No Addftioa Measures core barrel uppEr flang weld Primary baffle former asembly (plate and bolt) Primary bolts Prifmary hold down spring Primary core support column (ira*cessible) Expansion fthe*a OshiM flexuve A9001)ar radial spq k"ys 00, 900, 1800, 270° No Additional Measures secondary core support aasembly No Additional Measures thermal shield flexures (5) Prir/
lowr core suppor stucte Expansion lower core suppq columns Expansion lower core plate access cover " Existing thermal shield welds No Additional Measures tie plates (upper and lower) No Additional Measures core barrel to core u forging weld Primary upport column and fastener No Additional Measures guide tube assembly and fasteners ' Primary (cards) and No Additional Meapures (Fastenms) upper core plate keyway clevis 00, go0, 180,270, No Additional Measures top hat alignment key!ays 750, 1654, 2559, 3450 No Addional Measures upper core plate fuel assembly guide pins upper core plate periphery Expansion upper support assembly Existing Note: 1. Split pins are cndered to be a "fasieer n iWe with the guide tube assembly.
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United States Nuclear Regulatory Commission Enclosure to Serial: RNP-R/1 3-0073 Page 4 of 9 ONA 22: Cot*d RodE Gle Tube (CRQT)
RN 2-2:
a) Clarify whether the CRGT split pins are delined as ASME Section XI, Examination Category 8-W3 components for the HBRSEP CLB. If the CRGT split pins are defined in the ClB as ASME Section XI, Examination Category B44-3 core support structure components, justify why the CRGT split pins would not need to be inspected as part of the "Existing Program" or In accordance with an augmentation of the PWR Vessel Internals Program.
b) Justify why the response to ApplicaýtLicensee Action Item No. 3 did not specifically address whether the ASME Section Xl "Existing Program" bases in the CL0 for the CRGT split pins Is sufficient for ensuring adequate aging management of the split pins during the period of extended operation. Otherwise, provide a response to the Applicarnt/icensee Action Item No. 3 that specifically addresses the adequacy of ASME-Section XI "Existing Program" bases for the CRGT split pins and whether these bases are sufficient to manage aging in the components during the period of extended operation.
Rspense to OWI 2'2:
a) The CRGT split pins are defined as ASME Section X1, Examination Category B-N-3 components under the Robinson CLf[1, 2] as part of the guide tube assembly and fasterrs component. The CRGT split pins were not called out as an Existing component in MRP-227-A to support aging management MRP.227-A states that the guidance for guide tube support pins (split pins) is limited to plant-specific recommendations and subsequent performance monitoring should follow the supplier recommendations. Since Robinson has followed the Original Equipment Manufacturer (OEM) recommendations and replaced split pins they were categorized as a component under the "No Additional Measures" category for aging management considerations [7). MRP-227-A specifically states that there are no impacts to the plant-specific ASME Section Xl program but that only specific items that may be part of the program were credited as part of the internals aging management Inspection requirements. There is no impact to the ASME Section XI inspection requirements as a result of MRP-227-A implementation at Robinson.
b) The split pins are managed for aging under MRP-227-A requirements at Robinson as a result of the replacement performed compliant with OEM specifications; therefore consistent with OEM recommendations, the split pins are not included as a component requiring Inspection sampling as part of the MRP-227-A "Existing Program" component category. There is no change to the ASME Section Xl program at Robinson due to MRP-227-A Implementation and spVlt pins are included as part of the guide tube. Robinson is fully compliant with MRP-227-A and existing plant-specific ASME Section X1 requirements for split pins as detailed in plant-specific documents Identified in the aging management program plan (AMP).
Robinson has therefore demonstrated that aging for split pins are adequately managed during the period of extended operation.
United States Nuclear Re9tgatory Commission Enclosure to Serial: RNP-RA/13-0073 Page 6 of 9 RAM 2-3: KVI Ho-kDwn Spi for the RVI hold-a) Provide the besis for not idniyin the alternatve replacement basts measurement down spring as a deviation from the MRP-227-A recommended physical by the Electric basis, thus requirig the deviation to be reported and dispositond replacement is Power Research Institute (EPRI) MRP RVI Technical Committee. If a planned for the hold;-down spring, identify the limit on when the replacement activity wul need to be implemented and justify the timeframe selecte* for racemen of the component, b) Describe plans to manage aging of the replaced RI/I hold-down spring. If loss of preload due to stress relaxation or i o creep i still an aging effect that to Isapplicable to the replacement hold-down spdN, provide the basis for not needing spring implement the physical measurements in IARP-227-A to th replaced hold-downsprin, as well. if physical measurements woill be performed on 1he replaced hold-down provide and justify the acceptance criteria that will be applied to the physical No. 5.
measurement results, as recommended in Applicnt/Ltcensee Action Item Rso to RAI 2-3:
(May 21, 2013) that a) The industry has stated in recent public meetings with the regulator hold-down spring replacing the 304 stainless steel (SS) hold-down spring with a 403 SS is not a deviation from the MRP-227-A recommended physical measurement. The is not a component Identified for industry concluded that the 4035S hold-down spring The Robinson active aging management through the MRP-227-A a process.
in MRP-227-A, hold-down spring replacement will occur during the timeframe specified which is within three cycles of the beginning of the license renewal period.
to be b) With respect to stress relaxation, the behavior of Type 4035S wa considered certified 403 SS is a design significantly Improved over that of Type 304 SS. The Type degradation. In this alternative hold-down spring material and has no history of aging of the context, stress relaxation and creep defonnation are the inverse manifestations same physical effect. Type 403 SS is significantly more resistant to stress relaxation it displays lower creep rates under comparable creep than Type 304 stainless steel, as creep and stress conditions. in nuclear plants the thermal conditions, which drive although of deformation they do so relaxation, are insufficient to produce hgh rates relaxation under over tong service times. The rates of thermally driven creep and stress The rate of stress PWR conditions can be calculated from high temperature rates. persists even relaxation is lower for Type 403 SS than for Type 304 SS. This difference under condWits of irradiated enhanced creep/stress relaxation.
Thus, the Type 403 S can be expected to oontinue to provide less403 stress relaxation than would be observed InType 304 SS hold-down springs. Type SS is therefore in the hold-down springs expected to provide adequate resistance to stress relaxation Type 403 hold-for Robinson over the period of extended operation. On this basis, the V] and are down springs were screened out for stres relaxation effects in MRP-191 therefore not required to be inspected under the guidelines of MRP-227-A.
0
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/13-0073 Page 6 of 9 RAI 2-4: Operatng Experence RN 2-4:
a) Describe the programmatic activities that will be used to continually Identify plant-specific and industrial-aging issues. evaluate them, and, as necessary, enhance the PWR Vessel Internals Program. Identify whether these activities are consistent with the intent outlined in the NRC staffs latest guidance descbed in the Final License Renewal Interim Staff Guidance LR4SG-2011-05 "Ongoing review of Operating Experience." Otherwise, provide a basis for the conclusion that the current operating experience review activities for the AMP will be sufficient to ensure an adequate evaluation of operating experience on an ongoing basis and that the program will be appropriately adjusted, as necessary, to address age-related degradation inthe plant's RVW components during the period of extended operation.
b) Provide your basis why the "Operating Experince" program element does not provide a more comprehensive is of sources of relevant operating experience that will be reviewed by the licensee, Including operating experience that is discussed in the following sources: (i) Appendix A of MRP-227-A report (or in periok updates issued by the EPRI MRP); (ii)NRC issued generc communications, such as NUREG reports, information notices, generic letters, bulletins, or regulatory informaon summaries (11i) applica vendor reports, such as Westinghouse Commercial Atomic Power reports or vendor-issud degradation summaries; and (iv) PWR Owners Group repoft or EPRI MRP reports.
c) Provide the basis for not including inthe "Operating Experience' program element, nor evaluating the Impact on the PWR Vess Internals Program, the relevant operating experience associated with WestingWose-designed flux thimble tubes, thermashield/core barrel bolting, baffle-toformer bolts, clevis insert boIts/screws, and hold-down sp Rmpoome to Pl 2-4:
a) The Final License Renewal Interim Staff Guidance LR-ISG-201 1-05 (ADAMS ML12044A215)-Ongoing review of Operating Experience," recommends following the operating experience element of the GALL, which Section S.10 of WCAP-1 7077-NP 131 and Section 3.1.1.2 of the license renewal application (LRA) [6] are consistent with.
Robinson follows GALL, Revision 0, and as an enhancement, Robinson has reviewed the GALL as it was issued in Revislon 2, speifically for reactor vessel internals. As stated in Section 3.1.1.2 of the LRA [6], Robinson will continue to follow regulatory GALL updates or other pe t inrmatimon when issued, the industry Wrecommndations and opeai experience. As applicable, Robinson will continue th practice of enhancing plant-specific programs regularly as operan expeience is reported. Section 3.1.1.2 of the LRA [6] state inpart:
Operating experience (06)through December 2001 was considered during the development of the RNP integrated Plant Assessment. OE sWubeun to that date will be reviewed and applicable OE that requires any change to the CLB of Robinson
Uni States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA13-0073 Page 7 of 9 will be updated in conjunction with the amendment to the application required by 10 CFR 54.21(b).
The review consisted of the following:
Site: RNP site-specific operating experience was reviewed. The site-specific operating experience included a review of (1) Corrective Action Program, (2) Licensee Event Reports, (3) Maintenance Rule Data Base, and (4) interviews with Systems Engineers. No additional aging effects requiring management were identified beyond those Identified using the methods described in the previous Subsection.
Industry: An evaluation of industry operating experience published since the effective date of the GALL Report was performed to identify any additional aging effects requiring mnagement. No additional aging effects requiring management were Identified beyond those identified using the methods described In the previous Subsection.
On-Going: On-going review of plant-specific and industry operating experience is performed in accordance with the Corrective Action and Operating Experience Prografms.
b) In WCAP-17077-NP [3], under the "Operating Experience" program element, it is stated that industry operating experience is routinely reviewed by Duke Energy (Progress Energy) engineers using INPO Operating Experience (OE). the Nuclear Network, and other information sources as directed under the applicable procedure (CAP-NGGC-0202) [41, for the determination of additional actions and lessons learned. These insights, as applicable, are incorporated Into the plant systems quarterly health reports and further evaluated for Incorporation into plant programs.
WCAP-1 7077-NP also states, "Progress Energy (Duke Energy), through its participation in PWROG and EPRI-MRP activities, will continue to benefit from the reporting of Inspection information and will share its own operating experience with the industry through the reporting requirements of Section 7 of MRP-227-A."
Robinson has also made License Renewal Commitment 33 [51 which commits Robinson to:
- 1. Pwliar te in industryprogramsto investate aging effects and determinethe appropriateAMP activities to addressbaffle and former assembly issues, and to addresschanges in dimensions due to void swelling.
- 2. Evaluatethe results of completed researchprojects from the Westinghouse Owners Group (formerly hOG, now PWROG) and the EPRI MRP, and factor them into the PWR Vessel InternalsProgramas appropriate.
Robinson has demonstrated compliance with this commitment through continued participation as members of the MRP and PWROG including specialize committees related to reactor internals aging management and in the presentations made during the July 2012 IntearnationalBoit Water Reactor and Pressurized Water Reactor Materials Reliability Conference and Exhibition sponsored by the Electric Power Research Institute.
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/13-0073 Page 8 of 9 c) An extensive aging management review was completed in support of ilfe extension activities as documented in NGG Document, RNP-JLR-0354B, Rev. 2, [WCAP-17077-NP Reference 6]. The Robinson standard on-gOing practice Is to review technical bulletins and other industry information pertaining to internals components.
Updated assessments since completion of the life extension activities are summarized in the "Operating Experience" program element included as Section 5.10 in WCAP-17077-NP [3). This section specifically notes the proactiv aging activities done by Robinson with regards to the operating experience associated with the CRGT split pins and the CRGT guide cards. In addition Robinson active participation in industry activities was also noted and minutes of the April 2012 PWROG meeting included as WCAP-1 7077-NP Rigere 55. Although not specicy stated in the AMP these minutes provide details on recent relevant operating experience work1wide and supports the conclusion that there is no impact to report affecting Robinson as a result of recent operating experience in regards to thermal shield/core barrel bolting, baffleto-forier bolts, clvis Inert bolts/screws for currently operating PWRs of different designs, or the Combusin Engineering-design flux thimble tubes.
Robinson is following current industry guidanc and proactively replacing the hold-down spring with a Type 403 SS hold-ow spring, which was considered to be significantly improved over that of Type 304 8S. The Type 403 SS is a design certified alternative hold-down spring material and has no history of aging degradation.
United States Nuclear Regulatory Cormnission Erncosure to Sedal: RNP-RA/1 3-0073 Page 9 of 9 Rfeir sto 14 Respoau :
- 1. H.B. Robinson Document, RNP-PM-004, Rev. 17, "H. B. Robinson Steam Electric Plant, Unit No. 2, Fourth Ten-Year Interval, Inservice Inspection Plan."
- 2. H.B. Robinson Document, RNP-PM-009, Rev. 0, "H. B. Robinson Steam Electric Plant, Unit No. 2, Fifth Ten-Year Interval, Inservice Inspection Plan - Attachment
'B-N-i /B-N-2/B-N-3',"
- 3. Westinghouse Document, WCAP-17077-NP, Rev. 1, "PWR Vessel Internals Program Plan for Aging Management of Reactor Internal at Robinson Nuclear Plant," August 2012.
- 4. Nuclear Generation Group Standard Procedure, CAP-NGGC-0202, Rev. 22, "Operating Experience and Construction Experience Program."
- 5. U.S. Nuclear Regulatory Commission, NUREG- 1785, "Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2,"
Docket No. 50-261. Carolina Power & Light Company, March 2004.
- 6. Robinson Nuclear Plant Doment, RNP-RA/02-0089, "Application for Renewal of Operating License," June 14, 2002 (ADAMS ML021690663).
- 7. Materials Reliability Program: Screening, Categorization, and Rankang of Reactor InternalsComponentsfor Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.
- 8. Robinson Nuclear Plant Docmnt, RNP-RA/01 -0100, "Inservice Inspection Program for the Fourth Ten-Year Interval," August 17,2001 (ADAMS ML012330092).