RNP-RA/13-0053, Response to NRC Request for Additional Information Related to Pressurized Water Reactors Internals Program Plan for Aging Management of Reactor Internals
| ML13156A144 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/23/2013 |
| From: | Wheeler S Carolina Power & Light Co, Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/13-0053 | |
| Download: ML13156A144 (9) | |
Text
Sharon A. Wheeler-Peavyhouse H. B. Robinson Steam Electric Plant Unit 2 DUKE Manager -Support Services ENERGY Duke Energy Progress 8. T3581 West Entrance Road Hartsville, SC 29550 0: 843 857 1584 F: 843 857 1319 Sharon. WheeleKr4duke-energycon 10 CFR 54.21 Serial: RNP-RA/13-0053 MAY 2 3 2013 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 REACTORS INTERNALS PROGRAM PLAN FOR AGING MANAGEMENT OF REACTOR INTERNALS (TAC NO. ME9633)Ladies and Gentlemen:
By letter dated March 27, 2013, the NRC requested that Duke Energy Progress, Inc., formerly known as Carolina Power and Light Company, respond to a 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 the request for additional information is provided in the enclosure to this letter.There are no regulatory commitments made in this submittal.
If you should have any questions regarding this submittal, please contact Mr. R. Hightower at (843) 857-1329.I declare under penalty of perjury that the foregoing is true and correct.Executed On: M,1a- V-3 2- I-5 Sincerely, Sharon A. Wheeler-Peavyhouse Manager -Support Services -Nuclear United States Nuclear Regulatory Commission Serial: RNP-RA/13-0053 Page 2 of 2 SAWP/am
Enclosure:
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE PRESSURIZED WATER REACTORS INTERNALS PROGRAM PLAN FOR AGING MANAGEMENT OF REACTOR INTERNALS cc: Mr. V. M. McCree, NRC, Region II Ms. A. T. Billoch-Colon, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP Unit No. 2 United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RAI13-0053 Page 1 of 7 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE PRESSURIZED WATER REACTORS INTERNALS PROGRAM PLAN FOR AGING MANAGEMENT OF REACTOR INTERNALS (TAC NO. ME9633)DOCKET NO.: 50-261 NRC REQUEST FOR ADDITIONAL INFORMATION (RAI)By letter to the U.S. Nuclear Regulatory Commission (NRC) dated September 26, 2012, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12278A398), Duke Energy Progress, Inc., formerly known as Carolina Power and Light Company, submitted an aging management program for the reactor vessel internals for H. B.Robinson Steam Electric Plant, Unit No.2. The NRC staff has been reviewing the submittal and has determined that additional information is needed to complete its review.RAI-1 RAI-1: Historically, the following materials used in the PWR RVI components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A report. In this context, the NRC staff requests that the licensee confirm that these materials are not currently used in the RVI components at Robinson: 0 Nickel base alloys -Inconel 600; Weld Metals-Alloy 82 and 182 and Alloy X-750 (excluding control rod guide tube split pins)0 Alloy A-286 ASTM A 453 Grade 660, Condition A or B 0 Stainless steel type 347 material (excluding baffle-former bolts)0 Precipitation hardened stainless steel materials:
17-4 and 15-5 0 Type 431 stainless steel material Response to RAI-1: The Robinson pressurized water reactor (PWR) reactor vessel internals (RVI) components and their material of fabrication are confirmed in Westinghouse calculation CN-RIDA-12-34 Table 5-1 [1], which was transmitted to the NRC via [2].There are no components fabricated from the following materials at Robinson:-Alloy A-286 ASTM A 453 Grade 660, Condition A or B;-Precipitation hardened stainless steel materials:
17-4 and 15-5; or-Type 431 stainless steel material Robinson components fabricated from stainless steel type 347 material (excluding baffle- former bolts) include the following:
-Type 347 -barrel-former bolts Robinson components fabricated from the Nickel base alloys -InconelTM 600;Weld Metals-Alloy 82 and 182 and Alloy X-750 (excluding control rod guide tube split pins) include the following:
-Inconel 600 -flux thimble tube plugs-Inconel 600 -flux thimble tubes-Inconel 600 -clevis inserts-Inconel 600 -clevis insert lock keys United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA13-0053 Page 2 of 7-Alloy X-750 -clevis insert bolts-Alloy 82 weld metals -clevis insert lock keys are tack welded in place with Alloy 82; this is a non-structural weld.RAI-2 RAI-2: Condition 7 of the NRC staff's safety evaluation SE Revision 1, dated December 16, 2011, stipulates that the licensee shall include a summary of the operating experience related to the aging degradation in the RVI components.
The NRC staff requests that the licensee provide information regarding the extent of aging degradation (if any) that occurred thus far in all of the RVI components specifically, include the operating history of the following components at Robinson: Baffle-former bolts, baffle-edge bolts, baffle-former assembly, clevis insert bolts, core barrel bolting, and thermal shields.Provide a summary that includes a list of RVI components that have been inspected thus far, under the American Society of Mechanical Engineers Code,Section XI Inservice Inspection program and the inspection results. This list need not include any RVI component categorized under the "Existing" inspection category in the MRP-227-A report.Response to RAI-2: Information regarding aging degradation (if any) that occurred thus far in all of the Robinson RVI components has been summarized in several documents that have been provided previously or are available to the NRC, including:
0 WCAP-1 7077-NP, Revision 1, "PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant," Sections 4.2, 4.3.2.4, 5.10, and 6 [5]0 RNP-L/LR-0606, Revision 5,"Aging Management Program ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program" 0 RNP-L/LR-0354B, Revision 2, "Aging Management Review Reactor Internals" 0 MRP-219, Up to and Including Revision 8, "Materials Reliability Program: Inspection Data Survey Report" 0 WCAP-17435-NP, Revision 1, "Results of the Reactor Internals Operating Experience Survey Conducted under PWROG Project Authorization PA-MSC-0568" Section Xl inspections have been completed as required by the licensing requirements over the licensed life of the plant. A summary of all inspections completed under the Robinson ASME Code,Section XI Inservice Inspection Program (RNP-L/LR-0606) and the inspection results that include a list of RVI components that have been inspected thus far is contained in the plant records. There have been no incidents of degradation except as summarized in WCAP-17077-NP [5]. The operating and inspection history of the following components at Robinson is noted:
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RAI13-0053 Page 3 of 7 0 baffle-former bolts No abnormal conditions were observed.
MRP-227-A inspections will be completed in the fall 2013 outage (see WCAP-1 7077-NP, Section 7 [5])0 baffle-edge bolts A fully compliant MRP-227-A inspection was completed in the spring 2012 outage (see WCAP-1 7077-NP, Section 7 [5]). No degradation was reported.0 baffle-former assembly A fully compliant MRP-227-A inspection was completed in the spring 2012 outage (see WCAP-1 7077-NP, Section 7 [5]). No degradation was reported.0 clevis insert bolts ASME Code Section Xl inspections for the third period of the fourth inspection interval were completed in RO-27 spring 2012. No abnormal conditions were observed.core barrel bolting This is a MRP-227 "Expansion" item. This component will be inspected based on MRP-227-A Chapter 5 criteria and timing if degradation is observed, as detailed by MRP-227- A primary component inspections.
No degradation has been observed to date.thermal shields MRP-227 "Primary" component inspections for 100% of the thermal shield flexures compliant with MRP-228 requirements were completed in RO-27 spring 2012. No degradation was observed.RAI-3 RAI-3: According to Section A. 1.4 in MRP-175, "Materials Reliability Program: PWR Internal Aging Degradation Mechanism Screening Threshold Values," susceptibility to stress corrosion cracking (SCC) in nickel-based Alloy X-750 PWR RVI components depends on the type of heat treatment that is performed on the alloy. High temperature heat treatment (HTH)processes that are used on Alloy X-750 components offer better resistance to SCC than the other age hardened heat treatment processes.
Licensee determination of the heat treatment applied to its Alloy X-750 PWR RVI components would appear to be a critical parameter in ensuring the licensee's AMP will adequately manage the potential effects of aging.Additionally, Appendix A of the MRP-227- A report addressed, as a part of its operational experience, that Alloy X-750 used for the clevis insert bolt assembly in one unit failed due to pressurized water SCC (PWSCC). Therefore, the staff requests that the licensee provide information related to the type of heat treatment process that was used for the Alloy X-750 clevis insert bolting at Robinson.
If the existing clevis insert bolts at Robinson did not undergo an HTH process, the staff recommends that the licensee inspect these bolts (in addition to the inspections to monitor aging due to wear) for identifying PWSCC.Response to RAI-3: The Westinghouse documentation and material specification for the Alloy X-750 clevis insert bolting at Robinson confirms the X-750 bolt material was subjected to a single age-hardening step at a sustained temperature and subsequently air cooled [3]. The material United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/1 3-0053 Page 4 of 7 did not receive a HTH heat treatment.
Because the Alloy X-750 clevis insert bolts did not receive an HTH heat treatment, they might be susceptible to pressurized water stress corrosion cracking (PWSCC). However, the MRP-227-A [12] inspection management plan requires an ultrasonic examination of the baffle-former and barrel-former bolts to manage PWSCC in the reactor internals.
These bolts were identified as being more susceptible to PWSCC than the X-750 clevis insert bolts, as summarized in the MPR-191 ranking supporting the MRP-227-A inspection selections.
Robinson is following the MRP-227-A inspection requirements to support managing aging in the reactor internals.
RAI-4 RAI-4: In Appendix C, Table C-1 of the licensee's submittal dated September 26, 2012; the licensee indicated that the control rod guide tube cards are to be inspected no later than two refueling outages from the beginning of the period of extended operation.
This submittal stated that during the spring 2010 refueling outage the licensee performed inspections on control rod guide cards to assess the wear of these cards. The staff requests that the licensee provide the following information
-(1) the number of cards inspected; (2) the inspection results; (3) how the criteria for maximum allowed wear was established; (4)licensee's corrective actions, if any; and (5) the licensee's plan for subsequent inspections of this component during the extended period of operation.
Response to RAI-4: The Robinson upper internals guide tube-guide card wear evaluation is documented in WCAP-17277-P [4]. The following information, from [4], responds to the requests of RAI 4: (1) All nine guide cards, within nine of the guide tubes (20%, compliant with MRP-227-A requirements), were inspected at Robinson.(2) The inspection showed that the guide tubes can operate safely for the licensed life of the plant. Compliant with acceptance criteria, which is discussed in Section 2 of [4], no aggressive guide card inner wear or backside wear was identified at any of the Robinson measured guide tube locations.
No general trend in the ligament wear pattern was observed for any of the inspected guide tubes. Also, the ligament wear magnitude was extremely small.(3) The criteria for maximum allowed wear was established by Westinghouse Electric Company. The methodology for selecting guide card wear criteria is to determine the most credible mechanisms that could affect RCCA performance and then establish acceptance criteria that would allow continued operation until the next periodic inspection or guide tube replacement.
Various possible wear mechanisms were considered, and testing and analyses was performed to determine the possible degradation effect on RCCA performance, which is contained in Westinghouse Proprietary Letter LTR-RIDA-08-188, "15x15 Guide Card Wear Criteria Report." A brief description of the various analyses and testing performed to develop the wear criteria is presented below.Guide card drag test -The purpose of this test was to determine if the contact between a worn RCCA rodlet and a worn guide card could result in hanging up or catching of the rodlet during a controlled rod insertion or a gravity free-fall United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RAI13-0053 Page 5 of 7 Determination of collapse strength of a typical rodlet assembly -The purpose of this calculation note was to determine, by finite element (FE) simulation, the collapse strength of a typical rodlet assembly (rodlet host tube and coaxial internal absorber) due to axial (longitudinal) compressive forces that could occur during insertion or withdrawal of the RCCA Stress evaluation of the worn guide card holes -The purpose of the evaluation was to determine the limitation of the wear on guide card holes based on stresses remaining acceptable in the guide tube assembly Evaluation of RCCA rodlet wedging or sticking during or after a faulted event -The purpose of this evaluation was to determine a ligament thickness wear acceptance criterion for a 1 5x1 5 style guide tube that ensures that a rodlet will not wedge or stick during or after a faulted event In order to apply the acceptance criteria, it was determined that at the selected guide tubes, the holes with the smallest ligaments and expected aggressive wear should be visually examined and measured at every guide card span. There is concern for rodlet breakout; therefore, the slot width on the guide card is measured, and depending on the amount of wear, the guide card is placed in the green, yellow, or red zone for the acceptance criteria.(4) No corrective actions were required.(5) The plan for subsequent inspections of this component during the extended period of operation will follow the examination frequency, as identified in MRP-227-A.
As documented in WCAP-17077-NP
[5], the licensee will next inspect the guide cards in the Spring 2021 outage.RAI-5 RAI-5 (a): As discussed in Section 3.3.7 of the staffs SE for the MRP-227-A, Action Item 7 states that licensees of Westinghouse-designed reactors are required to develop plant-specific analyses to be applied to its facilities to demonstrate that lower support column cast austenitic stainless steel bodies will maintain their function during the extended period of operation.
This component is subject to neutron embrittlement and irradiation-assisted SCC (IASCC). Supplemental inspection would be recommended for those components that are potentially susceptible to neutron embrittlement, and are subject to significant tensile loadings under any normal operating or design basis condition.
Neutron embrittlement and IASCC become active aging degradation mechanisms when the fluence values exceed the threshold limit addressed in Table 4-6 of MRP-191 Revision 0, "Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs." The fluence value for the lower support column bodies in Westinghouse-designed reactors is significantly greater than the lx1 017 n/cm 2 threshold value provided in the Generic Aging Lessons Learned report. Therefore, the staff requests that the licensee submit an analysis taking into account the aforementioned aging effects. This analysis should consider the effects of transient loading on the structural integrity of the lower support column bodies.
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RN13-0053 Page 6 of 7 Response to RAI-5 (a): As summarized in MRP-1 91 [11], mechanical properties of some materials are altered by exposure to neutron irradiation or by long-term thermal aging. The fluence value, as summarized in MRP-1 91, affecting both irradiation embrittlement and irradiation-assisted stress corrosion cracking (IASCC), for the lower support column bodies in Westinghouse-designed reactors is significantly greater than the lx1 017 n/cm 2 threshold value provided in the Generic Aging Lessons Learned (GALL) report.WCAP-1 7077-NP Table 6-2 [5] provides a summary of the results of the evaluation
[6] of the Robinson cast austenitic stainless steel (CASS) material components, including that of the lower support column bodies. The evaluation was based on the NRC letter of May 19, 2000[9] and determined the components were not susceptible to thermal aging embrittlement.
Thus, the appropriate level for susceptibility to irradiation embrittlement is that cited in MRP-191, which is much higher than the lx1 017 n/cm 2 level provided in the GALL. Furthermore, WCAP-1 7254-P [7] contains Robinson plant-specific acceptance criteria, that demonstrates functionality of the component (i.e. the lower support column bodies), that is consistent with the WCAP-1 7096-NP [8] methodology.
RAI-5 (b): In Section 6.2.7 of the September 26, 2012, submittal, the licensee stated that the following components were evaluated for their susceptibility to thermal embrittlement based on the ferrite and Molybdenum contents and the casting process used. The staff requests that the licensee provide the methodology that was used to perform this evaluation (e.g., usage of an existing certified material test report). The components include: flow mixer devices; upper support column bases, with and without flow mixers; lower support columns; and bottom mounted instrumentation cruciforms, butt, and special columns.Response to RAI-5 (b): The Robinson reactor internal components that were manufactured from CASS are identified in WCAP-17077-NP Revision 1 Table 6-2 [5]. The components were evaluated for their susceptibility to thermal embrittlement based on the guidance of NRC letter of May 19, 2000 [9].As recommended in the NRC letter [9], Hull's Equivalent Factor equation in [10] was used to calculate delta ferrite content. Multiple sources were used to support the evaluation including certified material test reports (CMTRs) and material fabrication specifications containing specific chemical composition dictates.
Where chemistry data was not concise, molybdenum was conservatively assumed to be 0.5 percent (the maximum permitted for Grade CF8, per ASME requirements) and nitrogen (per guidance in [10]) was assumed to be 0.04 percent.
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/1 3-0053 Page 7 of 7 References to RAI Responses:
- 1. Westinghouse Calculation Note, CN-RIDA-12-34, Rev. 1, "H.B. Robinson Reactor Internals MRP-227-A Aging Management Program Plan Update Applicant/Licensee Action Items 1 and 2," August 22, 2012. (Westinghouse Proprietary Class 2).2. Westinghouse Affidavit, CAW-1 3-3618, "Application for Withholding Proprietary Information From Public Disclosure, PGN-12-71 Revision 1, "MRP-227-A Aging Management Program Plan Update" (Proprietary) and CN-RIDA-12-34 Revision 1,"H.B. Robinson Reactor Internals MRP-227-A Aging Management Program Plan Update Applicant/Licensee Action Items 1 and 2" (Proprietary)," February 8, 2013.3. Westinghouse Drawing, 206C017, Rev. 2, "PWR Clevis Insert .750 External Hex Cap Screw." 4. Westinghouse Report, WCAP-17277-P, Rev. 0, "H.B. Robinson Unit 2 -15x15 Upper Internals Guide Tube -Guide Card Wear Evaluation," October 2010.(Westinghouse Proprietary Class 2).5. Westinghouse Report, WCAP-1 7077-NP, Rev. 1, "PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant," August 2012.6. Westinghouse Calculation Note, CN-RIDA-12-42, Rev. 0, "H.B. Robinson Reactor Internals MRP-227-A Aging Management Program Plan Update Licensee Action Item 7, Evaluation of CASS," July 27, 2012. (Westinghouse Proprietary Class 2).7. Westinghouse Report, WCAP-17254-P, Rev. 0, "Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for H.B. Robinson Unit 2 Reactor Vessel Internals MRP-227 Primary and Expansion Components," August 2011. (Westinghouse Proprietary Class 2).8. Westinghouse Report, WCAP-1 7096-NP, Rev. 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," December 2009.9. NRC letter, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000 (NRC ADAMS Accession No. ML003717179).
- 10. U.S. Nuclear Regulatory Commission NUREG/CR-4513, Rev. 1, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," August 1994 (NRC ADAMS Accession No. ML052360554).
- 11. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191).
EPRI, Palo Alto, CA: 2006. 1013234.12. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A).
EPRI, Palo Alto, CA: 2011. 1022863.