L-PI-13-018, Materials Reliability Program (MRP) 227-A Applicant/Licensee Action Items Three Through Eight and Topical Report Conditions
| ML13067A284 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/07/2013 |
| From: | Jeffery Lynch Northern States Power Co, Xcel Energy |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-PI-13-018, MRP 227-A | |
| Download: ML13067A284 (10) | |
Text
tl Xcel Energy MAR 0 7 2013 L-PI-13-018 10 CFR 50.90 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 Materials Reliability Program (MRP) 227-A Applicant/Licensee Action Items Three Through Eight and Topical Report Conditions 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 the Prairie Island Nuclear Generating Plant(PINGP) PWR [Pressurized Water Reactor]
Vessel Internals Program which is based on Electric Power Research Institute Report 1016596, "Materials Reliability Program (MRP): Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" (MRP-227-A).
The NRC safety evaluation for MRP-227, Rev 0 (ML11308A770) contained eight action
,items for applicants/licensees to consider. In the October 1,2012 letter (ML12276A041)
NSPM stated that the considerations of these eight applicant/licensee action items would be provided to the NRC by April 1, 2013. By telephone call on February 5, 2013, the NRC Staff requested NSPM to provide considerations of MRP-227-A applicant/licensee action items three through eight prior to April 1, 2013 and also address seven topical report conditions discussed in the safety evaluation. NSPM considerations of MRP-227-A applicant/licensee action items three through eight and the topical report conditions are provided in the Enclosure to this letter.
If there are any questions or if additional information is needed, please contact Mr. Dale Vincent, P.E., at 651-388-1121.
1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Document Control Desk Page 2 Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
L~'i::!-
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
Enclosure Prairie Island Nuclear Generating Plant (PINGP)
Applicant/Licensee Action Items from NRC Safety Evaluation for Materials Reliability Program (MRP) 227, Rev 0 The PINGP PWR [Pressurized Water Reactor] Vessel Internals Program is based on Electric Power Research Institute Report 1016596, "Materials Reliability Program (MRP): Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" (MRP-227-A), which has been endorsed by the NRC (ML120270374).
Topical Report Conditions The NRC December 16, 2011 safety evaluation (SE) for MRP-227, Rev 0, (ML11308A770) included seven topical report conditions (TRC). By telephone call on February 5, 2013, the NRC Staff requested Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), to discuss the disposition of these TRCs. The NRC-approved PINGP Aging Management Review (AMR) Program and Aging Management Program (AMP) were evaluated for the impact of these TRCs and the results follow.
TRC 1: SE Section 4.1.1 High Consequence Components in the "No Additional Measures" Inspection Category For the PINGP, a Westinghouse design, this TRC changed the upper core plate and the lower core support forging from "no additional measures" components to the "Expansion" category, linked to the Control Rod Guide Tube (CRGT) lower flange weld "Primary" inspection item.
This TRC does not affect the Scoping and Screening because the upper core plate and the lower core support forging were already subject to AMR.
Based upon review of the NRC draft Interim Staff Guidance (ISG) LR-ISG-2011-04 in conjunction with this TRC, two changes to the AMR are required due to a change to the Aging Effects Requiring Management (AERM). The original PINGP AMR for the lower support forging did not identify cracking due to fatigue as an AERM. Also, the original AMR for the upper core plate did not include loss of fracture toughness due to neutron irradiation embrittlement, which was added by the ISG. The AMR will be revised to reflect these added AERM for these components.
The PWR Vessel Internals AMP will be credited with managing these newly identified AERM for the upper core plate and the lower support forging.
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Enclosure No revisions to the AMP are required because the AMP commits to implement MRP-227 -A which addresses TRC 1 by adding the upper core plate and lower core support forging to Table 4-6, Westinghouse Plants Expansion Components.
TRC 2: SE Section 4.1.2 Inspection of Components Subject to Irradiation-Assisted Stress Corrosion Cracking For PINGP, this TRC moved the "upper and lower core barrel cylinder girth welds" and the "lower core barrel flange weld" from the "Expansion" category to the "Primary" category, with the core barrel axial welds being the expansion link from these primary items.
This TRC does not affect Scoping and Screening because the core barrel and core barrel flange were already subject to AMR.
This TRC results in no effect on the existing PINGP AMR because the core barrel and core barrel flange were already associated with the AERM included in draft LR-ISG-2011-04 Table IV.B2. The original AMR considered dimensional change due to voids, irradiation-assisted stress corrosion cracking (IASCC), stress corrosion cracking (SCC),
loss of fracture toughness due to radiation embrittlement and void swelling, fatigue cracking, and loss of material due to wear as those AERM for the core barrel and core barrel flange component that credited the PWR Vessel Internals AMP for management.
No revisions to the AMP are required because the AMP commits to implement MRP-227-A which addresses TRC 2 by including the upper and lower core barrel cylinder girth welds and lower core barrel flange weld in Table 4-3, Westinghouse Plants Primary Components. "Upper and lower core barrel cylinder axial welds" are retained in Table 4-6, Westinghouse Plants Expansion Components, with a primary link from the upper and lower core barrel cylinder girth welds.
TRC 3: SE Section 4.1.3 Inspection of High Consequence Components Subject to Multiple Degradation Mechanisms This TRC does not affect the recommended program for Westinghouse-designed plants and therefore is not applicable to PINGP programs.
TRC 4: SE Section 4.1.4 Imposition of Minimum Examination Coverage Criteria for "Expansion" Inspection Category Components This TRC imposed a minimum coverage requirement on "Expansion" category components. At least 75% coverage must be obtained when the sample population is specified as 100%. When the specified population is given as "100% of accessible surfaces or components", then at least 75% of those accessible must be examined.
This TRC does not affect Scoping and Screening or AMR outputs because it is an inspection sample plan detail.
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Enclosure No revisions to the AMP are required because the AMP commits to implement MRP-227-A which addresses TRC 4 by including Note 2 in Table 4-6, Westinghouse Plants Expansion Components, which applies to: upper core plate; lower support forging; lower support column bolts; barrel-former bolts; core barrel outlet nozzle welds; upper and lower core barrel cylinder axial welds; core barrel outlet nozzle welds; and lower support column bodies (both cast and non-cast) inspection items.
TRC 5: SE Section 4.1.5 Examination Frequencies for Baffle-Former Bolts and Core Shroud Bolts This TRC specified a change in the examination interval for the baffle-former bolts for Westinghouse-designed plants from a range of 10-15 years to a more prescriptive fixed 10-year interval.
This TRC does not affect Scoping and Screening or AMR outputs because it is an inspection detail.
No revisions to the AMP are required because the AMP commits to implement MRP-227 -A which addresses TRC 5 by specifying the inspection frequency in Table 4-3, Westinghouse Plants Primary Components, as a fixed 10-year interval.
TRC 6: SE Section 4.1.6 Periodicity of the Re-examination of "Expansion" Inspection Category Components This TRC specified a fixed prescriptive 1O-year re-examination interval for "Expansion" category items once expansion has been triggered.
This TRC does not affect Scoping and Screening or AMR outputs because it is an inspection detail.
No revisions to the AMP are required because the AMP commits to implement MRP-227-A which addresses TRC 6 by specifying the re-examination interval for each "Expansion" category items in Table 4-6, Westinghouse Plants Expansion Components, as a fixed 1O-year interval.
TRC 7: SE Section 4.1.7 Updating of MRP-227 Appendix A This TRC required EPRI to revise Appendix A of MRP-227 to reference NUREG-1801, Revision 2, (GALL Rev. 2), XI.M16A for AMP guidance. Previously MRP-227 Rev.O, Appendix A, included content guidance for licensees to develop a license renewal (LR) ten element AMP. Upon the issuance of GALL Rev. 2, this guidance was no longer needed. The TRC also specified that MRP-227-A was to include an OE discussion in Appendix A.
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Enclosure This TRC requirement on EPRI was implemented by EPRI issuance of MRP-227-A.
The TRC does not require licensees to revise aging management programs or inspection plans.
AMR Changes In addition to the AMR changes noted above needed to address the Topical Report Conditions, the PINGP AMR was also reviewed against draft LR-ISG-2011-04 for all other reactor internals components. The following gaps were identified.
The original PINGP AMR did not identify cracking due to fatigue as an AERM for the clevis insert bolts. This aging effect was added in LR-ISG-2011-04, and is discussed in Note 2 to Table 4-9 in MRP-227-A. Therefore, cracking due to fatigue will be added to the AMR as an aging effect crediting the PWR Vessel Internals AMP.
The original PINGP AMR did not identify loss of fracture toughness due to neutron irradiation as an aging effect for the BMI columns and flux thimble guides component.
This AERM will be added to the AMR and the PWR Internals AMP credited for managing the effect.
The original PINGP AMR did not identify wear as an aging effect for the hold-down spring. This AERM will be added to the AMR and the PWR Internals AMP credited for managing the effect.
Applicant/Licensee Action Items The NRC safety evaluation for MRP-227, Rev 0 (ML11308A770) contained eight action items for applicants/licensees to consider. Paragraph 3.5.1 (2) of the SE (ML11308A770) states:
To ensure the MRP-227 program and the plant-specific action items will be carried out by applicants/licensees, applicants/licensees are to submit an inspection plan which addresses the identified plant-specific action items for staff review and approval consistent with the licensing basis for the plant. If an applicant/licensee plans to implement an AMP which deviates from the guidance provided in MRP-227, as approved by the NRC, the applicant/licensee shall identify where their program deviates from the recommendations of MRP-227, as approved by the NRC, and shall provide a justification for any deviation which includes a consideration of how the deviation affects both "Primary" and "Expansion" inspection category components.
By letter dated October 1, 2012 (ML12276A041), NSPM stated that considerations of these eight applicant/licensee action items (AILAI) would be provided to the NRC by April 1, 2013. By telephone call on February 5, 2013, the NRC Staff requested NSPM to provide considerations of AlLAls three through eight prior to April 1 ,2013. NSPM considerations of AlLAls three through eight are provided below.
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Enclosure A/LAI 3: SE Section 4.2.3 Evaluation of th~ Adequacy of Plant-Specific Existing Programs As addressed in Section 3.2.5.3 in this SE, applicants/licensees of CE
[Combustion Engineering] and Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an applicant's/Iicensee's existing programs, or to identify changes to the programs that should be implemented to manage the aging of these components for the period of extended operation. The results of this plant-specific analyses and a description of the plant-specific programs being relied on to manage aging of these components shall be submitted as part of the applicant's/Iicensee's AMP application. The CE and Westinghouse components identified for this type of plant-specific evaluation include: CE thermal shield positioning pins and CE in-core instrumentation thimble tubes (Section 4.3.2 in MRP-227), and Westinghouse guide tube support pins (split pins) (Section 4.3.3 in MRP-227).
This is Applicant/Licensee Action Item 3.
NSPM Response:
The AlLAI requires that plant-specific analyses justify the existing programs that are used to manage aging in the guide tube support pins (split pins) for Westinghouse plants. PINGP Units 1 and 2 are Westinghouse designed units, so the guide tube support pins (split pins) are addressed-as follows. According to MRP-191, only alloy X-750 guide tube support pins screen in as category 'C' components for which the aging effects are above the screening criteria. Guide tube support pins manufactured from 316 stainless steel (SS) are category 'A' components for which the aging effects are below the screening criteria. PINGP replaced the control rod guide tubes and split pins in 1986 to address the issue of alloy X-750 cracking. The redesigned guide tube support pins and split pins allowed the incorporation of several enhancements to mitigate to possibility of cracking. Many other plants have replaced their split pins.
However, by replacing the entire guide tube together with the split pins at PINGP, the modification was not constrained to use split pins of the same size, materials and design. The replacement split pins at PINGP incorporated a larger shank diameter, changed the material to 316 cold-worked (CW) SS, and included geometrical changes, such as the fillet radius. Type testing demonstrated a marked increase in fatigue life of the advanced 316 CW SS split pin compared to the original alloy X-750 split pin. Split pins manufactured from 316 SS are category 'A' per MRP-191, Table 7-2, which applies to "component items for which aging effects are below the screening criteria" and "aging degradation significance is minimal".
A/LAI 4: SE Section 4.2.4 B&W [Babcock and Wilcox] Core Support Structure Upper Flange Stress Relief As discussed in Section 3.2.5.4 of this SE, the B&W applicants/licensees shall confirm that the core support structure upper flange weld was stress relieved Page 5 of 8
Enclosure during the original fabrication of the Reactor Pressure Vessel in order to confirm the applicability of MRP-227, as approved by the NRC, to their facility. If the upper flange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component. If necessary, the examination methods and frequency for non-stress relieved B&W core support structure upper flange welds shall be consistent with the recommendations in MRP-227, as approved by the NRC, for the Westinghouse and CE upper core support barrel welds. The examination coverage for this B&W flange weld shall conform to the staff's imposed criteria as described in Sections 3.3.1 and 4.3.1 of this SE. The applicant's/Iicensee's resolution of this plant-specific action item shall be submitted to the NRC for review and approval. This is Applicant/Licensee Action Item 4.
NSPM Response:
Since PINGP Units 1 and 2 are Westinghouse designed units, this AlLAI is not applicable.
A/LAI 5: SE Section 4.2.5 Application of Physical Measurements as part of I&E
[Inspection and Evaluation] Guidelines for B&W, CE, and Westinghouse RVI
[Reactor Vessel Internals] Components As addressed in Section 3.3.5 in the SE, applicants/licensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-227 for loss of compressibility for Westinghouse hold down springs, and fo'r distortion in the gap between the top and bottom core shroud segments in CE units with core barrel shrouds assembled in two vertical sections. The applicant/licensee shall include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are consistent with the plants' licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 5.
NSPM Response:
PINGP Units 1 and 2 are Westinghouse designed units with hold-down springs manufactured from type 403 SS. MRP-227 -A provides requirements for type 304 SS but does not provide physical measurements for type 403 SS hold-down springs. Hold-down springs of Type 403 are Category 'A' per MRP-191 Table 7-2, which describes "those component items for which aging effects are below the screening criteria" and "aging degradation significance is minimal".
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Enclosure A/LAI 6: SE section 4.2.6 Evaluation of Inaccessible B&W Components As addressed in Section 3.3.6 in this SE, MRP-227 does not propose to inspect the following inaccessible components: the B&W core barrel cylinders (including vertical and circumferential seam welds), B&W former plates, B&Wexternal baffle-to-baffle bolts and their locking devices, B&W core barrel-to-former bolts and their locking devices, and B&W core barrel assembly internal baffle-to-baffle bolts. The MRP also identified that although the B&W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques. Applicants/licensees shall justify the acceptability of these components for continued operation through the period of extended operation by performing an evaluation, or by proposing a scheduled replacement of the components. As part of their application to implement the approved version of MRP-227, applicants/licensees shall provide their justification for the continued operability of each of the inaccessible components and, if necessary, provide their plan for the replacement of the components for NRC review and approval. This is Applicant/Licensee Action Item 6.
NSPM Response:
Since PINGP Units 1 and 2 are Westinghouse designed units, this AlLAI is not applicable.
A/LAI7: SE Section 4.2.7 Plant-Specific Evaluation of CASS [cast austenitic stainless steel] Materials As discussed in Section 3.3.7 of the SE, the applicants/licensees of B&W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B&W IMI [in-core monitoring instrumentation] guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during th!3 period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel or precipitation hardened stainless steel materials. These analyses shall also consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiring aging management during development of MRP-227. That is, the requirement would apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all Page 7 of 8
Enclosure licensing basis conditions of operation. The applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 7.
NSPM Response:
Section 3.3.7 of the SE (ML 1130SA770) states that the applicants/licensees of Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate Westinghouse lower support column bodies will maintain their functionality during the period of extended operation. AlLAI 7 further states that the requirement may not apply to components that were previously evaluated as not requiring aging management during development of MRP-227. The lower support columns for the internals of PINGP Units 1 and 2 are fabricated from wrought stainless steel rather than CASS, so the requirement for a plant-specific evaluation is not applicable. The bottom mounted instrumentation (8MI) cruciforms are comprised of CASS, as indicated in MRP-227, but these items have been classified as "no additional measures" components (final grouping of "N" in MRP-227 Table 3-3). No other CASS components were identified in the PINGP reactor internals as components requiring aging management.
A/LAI 8: SE Section 4.2.8 Submittal of Information for Staff Review and Approval As addressed in Section 3.5.1 in the SE, applicants/licensees shall make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by the SE, as an AMP for the RVI components at their facility.
This submittal shall include the information identified in Section 3.5.1 of the SE.
This is Applicant/Licensee Action Item S.
NSPM Response:
Section 3.5.1 of the SE (ML 1130SA770) states that applicants/licensees shall make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by the SE, as an AMP for the RVI components at their facility.
The PINGP application for license renewal (MLOS1130666) was submitted April 11, 200S, prior to the December 16, 2011 issuance of the SE to MRP-227 (ML 1130SA770).
The AMP for PINGP Units 1 and 2, containing the required ten elements and complying with MRP-227-A, was submitted for NRC review and approval by letter dated May 12, 2009 (ML091470263). The PINGP renewed operating licenses were issued on June 27,2011 (ML11147A140).
Due to an applicability review performed to satisfy AlLAI 1, NSPM has identified changes to the inspection plan submitted by letter dated October 1, 2012 (ML12276A041). These changes will be addressed with the response to AILA I 1 to be provided by April 1, 2013.
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