L-PI-13-045, Supplement to Prairie Island Nuclear Generating Plant (PINGP) Aging Management Program Submittals
| ML13175A333 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/24/2013 |
| From: | Jeffery Lynch Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-13-045, TAC MF0052, TAC MF0053 | |
| Download: ML13175A333 (7) | |
Text
Xcel Energy JUN 24 2013 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 L-PI-13-045 Supplement to Prairie Island Nuclear Generating Plant (PINGP) Aging Management Program Submittals (TAC Nos. MF0052 and MF0053)
By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12276A041), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), submitted an aging management program (AMP) for the reactor vessel internals (RVI) at PINGP, Units 1 and 2. Letters dated March 7, 2013 (ML13067A284) and March 21, 2013 (ML13084A378) provided additional AMP information. The Enclosure to this letter provides responses to NRC requests for additional information (RAls) in a letter dated May 24,2013 (ML13130A144).
If there are any questions or if additional information is needed, please contact Mr. Dale Vincent, P.E., at 651-388-1121.
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
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Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosures (1) cc:
Administrator, Region III, USNRC Project Manager, PINGP, USNRC Resident Inspector, PINGP, USNRC 1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Enclosure Supplement to Prairie Island Nuclear Generating Plant (PINGP) Aging Management Program Submittals (TAC Nos. MF0052 and MF0053)
By letter dated May 24,2013 (ML13130A144), the NRC requested additional information on the PINGP Reactor Vessel Internals (RVI) Aging Management Program described in submittals dated October 1,2012 (ML12276A041), March 7, 2013 (ML13067A284), and March 22, 2013 (ML13084A378). Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"),
response to the NRC requests for additional information (RAls) are provided as follows.
NRC RAI-1:
Historically, the following materials used in the PWR [Pressurized Water Reactor] 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 PINGP, Units 1 and 2.
(1) Nickel base alloys - Inconel600; Weld Metals - Alloy 82 and 182 and Alloy X-750 (excluding control rod guide tube split pins)
(2) Alloy A-286 ASTM A 453 Grade 660, Condition A or B (3) Stainless steel type 347 material (excluding baffle-former bolts)
(4) Precipitation hardened (PH) stainless steel materials 4 and 15-5 (5) Type 431 stainless steel material NSPM response:
(1) Nickel base alloy is used in the clevis inserts, clevis insert bolts, and clevis insert bolt locking bars.
(2) Alloy A-286 ASTM A 453 Grade 660, Condition A or B, is not used in the PINGP reactor internals.
(3) Stainless steel type 347 is used in the barrel-former bolts, in addition to baffle-former bolts.
(4) Precipitation hardened (PH) stainless steels 17-4 and 15-5 are not used in the PINGP reactor internals.
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Enclosure (5) Type 431 stainless steel is not used in the PINGP reactor internals.
NRC RAI-2:
Condition 7 of Revision 1 of the NRC staff's December 16, 2011, SE [safety evaluation],
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 has occurred thus far in all of the RVI components. Specifically, include the operating history of the following components at PINGP, Units 1 and 2:
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 (ASME) Code,Section XI, inservice inspection program, and the inspection results. This list shall include any RVI component categorized under the "Existing" inspection category in the MRP-227-A report.
NSPM response:
ASME Section XI in-service examinations for category 8-N-3 have been performed once for each of the 1 S\\ 2nd, and 3rd 1 O-year intervals for both PINGP units. There has been no aging degradation identified to date with the PINGP reactor internals, including the baffle-former bolts, baffle-edge bolts, baffle-former assembly, clevis insert bolts, core barrel bolting, and thermal shields.
The following internals components have been inspected for ASME Section XI examinations for category 8-N-3 to date:
Thermocouple conduit, connections, and support brackets Thermocouple columns Thermocouple and flow mixer devices Upper core plate keyways Control rod guide tubes (upper)
Control rod guide tubes (lower)
Control rod guide tube fasteners and support pins Upper support columns Upper support column fasteners Page 2 of 6
Enclosure Upper core plate fuel assembly guide pins Upper core plate Upper core plate alignment keys Upper support weldment Upper support weldment keyways for head and vessel alignment pins Head and vessel alignment pins Core barrel outlet nozzle Irradiation specimen baskets Irradiation specimen guides and welds Specimen port plugs Access plate (also known as access port plug)
Instrumentation guides, collars, welds, and locking devices Lower core plate Lower core plate fuel assembly alignment pins Hold down spring Thermal shield Thermal shield bolts and pins Thermal shield flexures Thermal shield, flexure, welds, bolts, and pins Radial support keys Radial support key welds and fasteners Lower support column Core support structure Diffuser plate Tie plate Secondary core support assembly Energy absorber Clevis inserts (radial support keyway)
Clevis insert welds and fasteners Core barrel midplane weld Core barrel flange weld Safety injection nozzle interface Baffle assembly fasteners at top former Core barrel flange (upper internals support ledge)
Core barrel flange head cooling flow nozzle welds The MRP-227 -A "Existing Programs" Items that are credited to be managed for aging by the ASME Section XI examinations for Westinghouse plants are:
Core barrel flange Upper support ring or skirt (part of upper support weldment)
Lower core plate Clevis insert bolts Upper core plate alignment pins Page 3 of 6
Enclosure Each of the MRP-227 -A "Existing Programs" items above has been visually examined (VT-3) under the ASME Section XI In-service Inspection Program.
There have been no relevant indications identified with any of the MRP-227 -A "Existing Programs" items in any of the ASME Section XI in-service inspections for category B-N-
- 3.
NRC RAI-3:
According to Section A.1.4 in MRP-175, "Materials Reliability Program: PWR Internal Aging Degradation Mechanism Screening Threshold Values," the 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. Additionally, Appendix A of the MRP-227 -A report identified, as a part of the industry's operational experience, that the clevis insert assembly in Alloy X-750 bolting in one operating unit failed due to primary water stress corrosion cracking (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 materials used in the RVI components at PINGP. If Alloy X-750 material is used for clevis insert bolting or for any other RVI components at PINGP, Units 1 and 2, confirm that HTH treatment was performed on this material. If the clevis insert bolting did not undergo HTH treatment, discuss your plans to inspect these bolts (in addition to the inspections to monitor aging due to wear) for identifying PWSCC.
NSPM response:
Records of the heat treatment condition for the clevis insert bolts have not been retrieved, so the clevis insert bolts are conservatively assumed to be of the more susceptible AH or BH condition, which were common at the time of PINGP construction.
The clevis insert bolts were screened in for SCC in Electric Power Research Institute technical report MRP-191 and were classified as Category 'B'. The aging effect is managed under Existing Programs by the ASME Section XI In-Service Inspection Program. The clevis insert bolts (lower radial support keyway) are visually examined with the VT-3 method once per code interval. In the industry operating experience of failures with the clevis insert bolts, the failed bolts were readily detected visually. The bolt heads, while they remained captured by the locking bars in their countersunk recesses, were free to vibrate in the flow stream and had begun to show obvious wear to the locking bars and the corresponding slots in the bolt head. The bolt heads were visually cocked in their recesses, which was a clear indication of separation. The failed bolts in the industry operating experience did not lead to a loss of function of the lower radial support. Thus, the visual inspections under the existing ASME Section XI lSI program were sufficient to manage the aging effect.
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Enclosure NRC RAI-4:
In Enclosure 1, Page 2, of the licensee's October 1, 2012, submittal, the licensee indicated that the control rod guide tube (CRGT) cards will be inspected no later than two refueling outages from the beginning of the license renewal period (i.e., the period of extended operation). The licensee also stated that it would perform inspections on the CRGT cards to assess the wear of these cards. The staff requests that the licensee provide the following information:
(1) The number of cards that are planned to be inspected (2) The inspection results (3) How the criteria for maximum allowed wear was established (4) The licensee's corrective actions, if any, and (5) The licensee's plan for subsequent inspections of this component during the period of extended operation.
NSPM response:
(1) NSPM intends to comply with the requirements of WCAP-17451, "Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections", currently under final review with the PWR Owners Group Materials Subcommittee. Draft WCAP-17451 requires that the lower six guide cards above the continuous guidance section within a given guide tube be examined. Each PINGP unit has 29 control rod guide tubes. Both MRP-227-A and WCAP-17451 require a 20% sample. WCAP-17451 provides guidance for scope expansion based on the measurements of the 20% initial sample.
(2) No inspections of this type have been performed to date on the PINGP replacement upper internals, so no inspection results are currently available.
(3) Maximum acceptable wear criteria is based on preventing the rod cluster control assembly (RCCA) rod lets from wearing the guide card slots open to the extent that the slot width is equal to or greater than the rodlet diameter, at which time the guide card no longer supports the rod let from buckling due to LOCA crossflow forces or on a reactor trip when the rods are dropped. Detailed analyses within WCAP-17451 determine the number of consecutive guide cards that must be worn through for a rod let to have a potential for plastic deformation due to LOCA crossflow forces or buckling during insertion. In all cases (for different CRGT designs) there must be more than one consecutive guide card ligament completely worn through for there to be a loss of guidance. The acceptance limits in WCAP-17451 include an allowance for further wear prior to the next inspection. The wear rate projections are based on detailed finite element analyses considering static forces and flow-induced vibration for each of several RCCA guide tube designs, and the predicted wear rates have been successfully benchmarked against plant field results.
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Enclosure (4) There are currently no corrective actions in place or planned for this phenomenon. Once the initial round of inspections is complete at PINGP, corrective actions may be required based on the degree of wear identified.
These criteria are also contained in WCAP-17451.
(5) MRP-227 -A requires that the guide cards be reexamined on a 1 O-year interval.
This interval is specified within the supplemental information submitted under letter dated March 22, 2013 (ML13084A378).
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