L-PI-13-024, Materials Reliability Program (MRP) 227-A Applicant/Licensee Action Items One and Two

Materials Reliability Program (MRP) 227-A Applicant/Licensee Action Items One and Two

ML13084A378
Person / Time
Site:	Prairie Island
Issue date:	03/22/2013
From:	Jeffery Lynch Northern States Power Co, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-PI-13-024
Download: ML13084A378 (23)

Text

tl Xcel Energy MAR 2 2 2013 L-PI-13-024 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 One and Two 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 applicantsllicensees to consider. In the October 1, 2012 letter (ML12276A041)

NSPM stated that the considerations of these eight applicant/licensee action items as applicable to PINGP would be provided to the NRC by April 1, 2013. By letter dated March 7, 2013 (ML13067A284) NSPM submitted considerations of MRP-227-A applicant/licensee action items three through eight. NSPM considerations of MRP-227-A applicant/licensee action items one and two as applicable to PINGP 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.

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Site Vice¥r:s~~ent, 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).

Applicant/Licensee Action Items The NRC safety evaluation (SE) 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 [Aging Management Program]

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), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM") stated that considerations of these eight applicant/licensee action items (AILAI) as applicable to PINGP would be provided to the NRC by April 1,2013. By letter dated March 7, 2013 (ML13067A284) NSPM submitted considerations of MRP-227-A applicant/licensee action items three through eight. NSPM considerations of MRP-227-A applicant/licensee action items one and two as applicable to PINGP are provided below.

A/LAI 1: SE Section 4.2.1 Applicability of FMECA [Failure Modes, Effects, and Criticality Analyses] and Functionality Analysis Assumptions As addressed in Section 3.2.5.1 of the SE, each applicant/licensee is responsible for assessing its plant's design and operating history and demonstrating that the approved version of MRP-227 is applicable to the facility. Each Page 1 of 5

Enclosure applicant/licensee shall refer, in particular, to the assumptions regarding plant design and operating history made in the FMECA and functionality analyses for reactors of their design (Le., Westinghouse, CE [Combustion Engineering], or B&W [Babcock & Wilcox]) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI [Reactor Vessel Internal] components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227. This is Applicant/Licensee Action Item 1.

NSPM Response:

MRP-227-A in Section 2.4 provided the following criteria for determining its applicability to a nuclear facility:

1. No more than 30 years of high-leakage core designs, beyond which low-leakage core designs are used.

2. Base load operation

3. No design changes beyond those identified in general industry guidance or recommended by the original vendors.

Westinghouse assessed applicability criterion 1. above relative to PINGP and concluded the low-leakage core design criterion was satisfied, since PINGP transitioned to low-leakage cores early in plant life.

PINGP has been base-load operated through the majority of its life, but there were periods of load-following operation conducted in the 1970s that required consideration.

The effect of load-following on the limiting fatigue item in the reactor internals, the baffle-former bolts (BFB), was analyzed. The increased cycles due to the limited history of load-following do not result in a cumulative usage factor greater than one for the most limiting BFB for 60 years of operation and, therefore, the base load operation criterion is satisfied.

The Prairie Island Upper Internals Replacement Project was considered under the "no design changes" criterion since this component replacement could affect some of the assumed relationships between components within the internals used in the development of the MRP-227 aging management strategies. The complete upper internals assemblies for both PINGP Units 1 and 2 were replaced in 1986 after approximately 13 and 12 years of operation, respectively, making the replacement upper internals somewhat younger compared to the original lower internals which remain in service.

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Enclosure An expert panel process similar to that used to develop the aging management strategy contained in MRP-227-A was employed to evaluate the impact of the upper internals replacement and determine a technically justified alternative, if necessary. MRP-227-A Section 2.4 states, "These guidelines are also considered applicable to plants that have replaced components or component assemblies; however, alternatives can be technically justified."

The components of the upper internals which are used by MRP-227 -A as the leading indicators to manage aging effects are the CRGT lower flange welds (SCC, fatigue, thermal embrittlement, and irradiation embrittlement) and the CRGT guide cards (wear).

The control rod guide tube (CRGT) guide cards have no expansion link, so their reduced time-in-service does not affect the aging management strategy. The "Existing" program item "core barrel flange" is also credited with managing wear and may be treated as the alternative leading component for this aging effects requiring management (AERM), according to the expert panel results.

The CRGT lower flange welds are Primary inspection items that expand to the bottom mounted instrumentation (BMI) column bodies, lower support column bodies (cast),

upper core plate, and the lower support forging/casting.

Because the upper core plate was also replaced as part of the Prairie Island Upper Internals Replacement Project, the relationship to the CRGT lower flange welds is not changed compared to that assumed in the development of the aging management strategy (that is, they are the same age).

PINGP has non-cast lower support columns, so this expansion item is not applicable and the upper internals replacement has no relevancy to this item.

The CRGT lower flange welds are the Primary inspection item expanding to the BMI column bodies and the lower support forging. The expert panel process selected an alternate Primary link from the lower internals that would be a suitable lead indicator for the aging mechanisms of stress corrosion cracking (SCC) and fatigue, and concluded that the lower core barrel flange weld possessed the necessary attributes to be the correct leading indicator of SCC and fatigue for these Expansion components. An expansion link to the BMI column bodies and the lower support forging was added from the lower core barrel flange weld, in addition to the link from the CRGT lower flange weld, which was retained as a Primary inspection component.

Note that once implemented, the changes to the Primary-to-Expansion component relationships only adds requirements to the MRP-227-A inspections: none are removed.

Therefore, MRP-227-A is applicable to PINGP and the question created by the reduced age of the PINGP upper Internals assemblies is addressed conservatively through compensatory additional measures.

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Enclosure In addition to the minor strategy changes induced by the Prairie Island Upper Internals Replacement Project, Westinghouse evaluated whether the PINGP internals were well-represented by the MRP-191 Screening and Risk Ranking and MRP-232 Functionality Analysis processes with respect to the temperature, fluence, and stress assumptions used. The Westinghouse report concluded that PINGP reactor internals were well-represented.

Westinghouse reviewed the changes to the inspection plan necessitated by the Prairie Island Upper Internals Replacement Project impacts on the Primary-to-Expansion relationships and concluded that the Scoping and Screening, and Aging Management Review (AMR) documents were not affected. The Aging Management Program (AMP) was modified to include the new Primary-to-Expansion component relationships.

Revised inspection plan tables incorporating the additional expansion links are included as Attachment 1 to this enclosure. These tables supersede in their entirety those previously provided by letter dated October 1,2012 (ML12276A041).

A/LAI 2: SE Section 4.2.2 PWR Vessel Internal Components Within the Scope of License Renewal As discussed in Section 3.2.5.2 of the SE, consistent with the requirements addressed in 10 CFR 54.4, each applicant/licensee is responsible for identifying which RVI components are within the scope of LR [license renewal] for its facility.

Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the RVI components that are within the scope of LR for their facilities in accordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) and propose any necessary modifications to the program defined in MRP-227, as modified by the SE, when submitting its plant-specific AMP. The AMP shall provide assurance that the effects of aging on the missing component(s) will be managed for the period of extended operation. This issue is Applicant/Licensee Action Item 2.

NSPM Response:

The components in the PINGP Unit 1 and 2 reactor internals that were within the scope of license renewal were reviewed to ensure inclusion in Table 4-4 in MRP-191. The actual materials of construction used in the plant match those evaluated in MRP-191 for each component, or the differences were inconsequential (e.g., 316 stainless steel (SS) used instead of 304 SS assumed in MRP-191)

Because no mismatch was discovered between the scope and materials considered in MRP-191 and that evaluated in PINGP's Scoping and Screening, AMR and AMP, no changes were determined to be necessary.

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Enclosure Pursuant to SE Section 3.5.1 Item (2), the inspection plan complying with MRP-227-A is provided as Attachment 1 to this enclosure. This submittal modifies the inspection plan previously submitted by letter dated October 1, 2012 (ML12276A041) as follows. Due to the applicability review performed to satisfy AlLAI 1 as discussed above, an expert panel recommended additional primary-to-expansion links from the lower core barrel flange weld to the BMI column bodies and the lower support forging. These have been incorporated in the revised inspection plan included as Attachment 1 to this enclosure.

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Attachment 1 to Enclosure to letter L-PI-13-024 Prairie Island Nuclear Generating Plant (PINGP)

Applicant/Licensee Action Items from NRC Safety Evaluation for Materials Reliability Program (MRP) 227, Rev 0 15 pages follow

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 39 of 53 Attachment 1 MRP-227-A Primary Expansion, and Existing Programs Inspection Requirements Primary Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Control Rod Guide All Plants Loss of Material (Wear) None Visual (VT-3) examination no later 20% examination of the number Tube Assembly than 2 refueling outages from the of CRGT assemblies, with all Guide plates (cards) beginning of the license renewal guide cards within each period, and no earlier than two selected CRGT assembly refueling outages prior to the start of examined.

the license renewal period.

Subsequent examinations are required on a ten-year interval.

Control Rod Guide All Plants Cracking (SCC, Fatigue) Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible)

Tube Assembly Aging instrumentation examination to determine the CRGT lower flange weld Lower flange welds Management (IE and (BMI) column bodies, presence of crack-like surface flaws surfaces and adjacent base TE) Lower support in flange welds no later than 2 metal on the individual column bodies (cast, Not refueling outages from the beginning periphery CRGT assemblies.

Applicable to PINGP of the license renewal period and (Note 2).

1&2) subsequent examination on a ten-Upper core plate year interval.

Lower support forging/casting Core Barrel Assembly All Plants Cracking (SCC) Lower Support Periodic enhanced visual (EVT-1) 100% of one side of the Upper core barrel flange column bodies (non examination, no later than 2 refueling accessible surfaces of the weld cast) outages from the beginning of the selected weld and adjacent Core barrel outlet license renewal period and base metal (Note 4).

nozzle welds. subsequent examination on a ten-year interval.

Core Barrel Assembly All Plants Cracking (SCC, IASCC, Upper and lower core Periodic enhanced visual (EVT-1) 100% of one side of the Upper and lower core Fatigue) barrel cylinder axial examination, no later than 2 refueling accessible surfaces of the barrel cylinder girth welds outages from the beginning of the selected weld and adjacent welds license renewal period and base metal (Note 4).

subsequent examination on a ten-year interval.

. L.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

REACTOR VESSEL INTEGRITY H44 H PROGRAM REV: 15 Page 40 of 53 Attachment 1 MRP-227 -A Primary Expansion, and Existing Programs Inspection Requirements Primary Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Core Barrel Assembly All plants Cracking (SCC, Fatigue) Bottom-mounted Periodic enhanced visual (EVT-1) 100% of one side of the Lower core barrel flange instrumentation (BMI) examination, no later than 2 refueling accessible surfaces of the weld (Note 5) column bodies, outages from the beginning of the selected weld and adjacent Lower support license renewal period and base metal (Note 4).

forging/casting subsequent examination on a ten-year interval.

Baffle-Former All plants with Cracking (IASCC, None Visual (VT-3) examination, with Bolts and locking devices on Assembly baffle-edge bolts Fatigue) that results in: baseline examination between 20 high f1uence seams. 100% of Baffle-edge bolts

• Lost or broken and 40 EFPY and subsequent components accessible from locking devices examinations and a ten-year interval. core side (Note 3).

• Failed or missing bolts

• Protrusion of bolt heads Aging Management (IE and ISR) (Note 6)

Baffle-Former All plants Cracking (IASCC, Lower support column Baseline volumetric (UT) 100% of accessible bolts (Note Assembly Fatigue) Aging bolts, Barrel-former bolts examination between 25 and 35 3). Heads accessible from the Baffle-former bolts Management (IE and EFPY, with subsequent examination core side. UT accessibility may ISR) (Note 6) on a ten-year interval. be affected by complexity of head and locking device designs.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 41 of 53

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Attachment 1 MRP-227 -A Primary Expansion, and Existing Programs Inspection Requirements Primary Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Baffle-Former All plants Distortion (Void None Visual ((VT-3) examination to check Core side surface as indicated.

Assembly Swelling), or Cracking for evidence of distortion, with Assembly (lASCC) that results in: baseline examination between 20 (includes: Baffle plates, -Abnormal and 40 EFPY and subsequent baffle edge bolts and interaction with examinations on a ten-year interval.

indirect effects of void fuel assem blies swelling in former plates) -Gaps along high fluence baffle joint

-Vertical displacement of baffle plates near high fluence joint

-Broken or damaged edge bolt locking systems along high fluence baffle joint Alignment and All plants with 304 Distortion (Loss of load) None Direct measurement of spring height Measurements should be taken Interfacing stainless steel within three cycles of the beginning at several points around the components hold down springs Note: This mechanism of the license renewal period. If the circumference of the spring, Internals hold down was not strictly identified first set of measurements is not with a statistically adequate spring Not Applicable to in the original 'list of sufficient to determine life, spring number of measurements at PINGP 1&2 age-related degradation height measurements must be taken each point to minimize mechanisms [7].' during the next two outages, in order uncertainty.

to extrapolate the expected spring height to 60 years.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 42 of 53 Attachment 1 MRP-227 -A Primary Expansion, and Existing Programs Inspection Requirements Primary Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Thermal Shield All plants with Cracking (Fatigue) or None Visual (VT-3) no later than 2 100% of thermal shield flexures.

Assembly thermal shields Loss of Material 0Near) refueling outages from the beginning Thermal Shield that results in thermal of the license renewal period.

Assembly shield flexures excessive Subsequent examinations on a ten-wear, fracture, or year interval.

complete separation.

Notes to Primary ComponentsTable:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Attachment 2.

2. A minimum of 75% of the total identified sample population must be examined.

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Attachment 2, must be examined for inspection credit.

4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Attachment 2, must be examined from either the inner or outer diam eter for inspection credit.

5. The lower core barrel flange weld may be alternatively designated as the core barrel-ta-support plate weld in some Westinghouse plant designs.

6. Void swelling effects on this component is managed through management of void swelling on the entire baffle-former assembly.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 43 of 53 Attachment 1 MRP-227-A Primary Expansion, and Existing Programs Inspection Requirements Expansion Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Upper Internals All plants Cracking CRGT lower flange weld Enhanced visual (EVT-1) 100% of accessible surfaces Assembly (Fatigue, Wear) examination. (Note 2).

Upper core plate Re-inspection every 10 years following initial inspection.

Lower Internals All plants Cracking CRGT lower flange weld Enhanced visual (EVT-1) 100% of accessible surfaces Assembly Aging management (TE and Lower core barrel examination. (Note 2).

Lower support forging or in Casting) flange weld Re-inspection every 10 years castings following initial inspection.

Core Barrel Assembly All plants Cracking Baffle-former bolts Volumetric (UT) examination. 100% of accessible bolts.

Barrel-former bolts (IASCC, Fatigue) Re-inspection every 10 years Accessibility may be limited by Aging Management (IE, following initial inspection. presence of thermal shields or Void Swelling and ISR) neutron pads (Note 2).

Lower Support All plants Cracking Baffle-former bolts Volumetric (UT) examination. 100% of accessible bolts or as Assembly (IASCC, Fatigue) Re-inspection every 10 years supported by plant-specific Lower support column Aging Management (IE following initial inspection. justification (Note 2).

bolts and ISR)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 44 of 53 Attachment 1 MRP-227-A Primary Expansion, and Existing Programs Inspection Requirements Expansion Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Core Barrel Assembly All plants Cracking (SCC, Fatigue) Upper core barrel flange Enhanced visual (EVT-1) 100% of one side of the Core barrel outlet nozzle Aging Management (IE weld examination. accessible surfaces of the welds of lower sections) Re-inspection every 10 years selected weld and adjacent following initial inspection. base metal (Note 2).

Core Barrel Assembly All plants Cracking (SCC, IASCC) Upper and lower core Enhanced visual (EVT-1) 100% of one side of the Upper and lower core Aging Management (IE) barrel cylinder girth examination. accessible surfaces of the barrel cylinder axial welds Re-inspection every 10 years selected weld and adjacent welds following initial inspection. base metal (Note 2).

Lower Support All plants Cracking (IASCC) Upper core barrel flange Enhanced visual (EVT-1) 100% of accessible surfaces Assembly Aging Management (IE) weld examination. (Note 2).

Lower support column Re-inspection every 10 years bodes (non cast) following initial inspection.

Lower Support All plants Cracking (lASCC) Control rod guide tube Visual (EVT-1) examination. 100% of accessible support Assembly including the detection of (CRGT) lower flanges Re-inspection every 10 years columns (Note 2).

Lower support column fractured support following initial inspection.

bodes (cast) Not Applicable to columns PINGP 1&2 Aging Management (IE)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 45 of 53 Attachment 1 MRP-227-A Primary Expansion, and Existing Programs Inspection Requirements Expansion Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Bottom Mounted All plants Cracking (Fatigue) Control rod guide tube Visual (VT-3) examination of BMI 100% of BMI column bodies for Instrumentation including the detection of (CRGT) lower flanges column bodies as indicated by which difficulty is detected System completely fractured and Lower core barrel difficulty of insertion/withdrawal of during flux thimble Bottom-mounted column flange weld flux thimbles. insertion/withdrawal.

instrumentation (BMI) Aging Management Re-inspection every 10 years column bodies (IE). following initial inspection.

Flux thimble insertion/withdrawal to be monitored at each inspection interval. -----

Notes to Expansion Components Table:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Attachment 2.

2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 46 of 53 Attachment 1 MRP-227 -A Primary Expansion, and Existing Programs Inspection Requirements Existing Program Components Expansion Link Examination Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Core Barrel Assembly All plants Loss of material (Wear) ASME Code Section XI Visual (VT-3) examination to All accessible surfaces at Core barrel flange determine general condition for specified frequency excessive wear.

Upper Internals Assembly All plants Cracking (SCC, Fatigue) ASME Code Section XI Visual (VT-3) examination. All accessible surfaces at Upper support ring or skirt specified frequency Lower Internals Assembly All plants Cracking (IASCC, ASME Code Section XI Visual (VT-3) examination of the All accessible surfaces at Lower core plate Fatigue) lower core plates to detect evidence specified frequency XL lower core plate (Note 1) Aging Management (IE) of distortion and/or loss of bolt integrity.

Lower Internals Assembly All plants Loss of material (Wear) ASME Code Section XI Visual (VT-3) examination All accessible surfaces at Lower core plate specified frequency XL lower core plate (Note 1)

Bottom Mounted All plants Loss of material (Wear) NUREG-1801 Surface (En examination. Eddy current surface Instrumentation System Rev. 1 examination as defined in Flux thimble tubes plant response to IEB 88-09.

Alignment and Interfacing All plants Loss of material (Wear) ASME Code Section XI Visual (VT-3) examination All accessible surfaces at Components speCified frequency Clevis insert bolts (Note 2)

Alignment and Interfacing All plants Loss of material (Wear) ASME Code Section XI Visual (VT-3) examination All accessible surfaces at Components specified frequency Upper core plate alignment pins Notes to Existing Programs Components Table:

=

1. XL "Extra Long" referring to Westinghouse plants with 14-foot cores.

2. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the c1evislinsert is the issue.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 47 of 53 Attachment 2 MRP-227 -A Examination, Acceptance and Expansion Criteria Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Link(s) Acceptance Criteria (Note 1)

Control Rod Guide Tube All plants Visual (VT-3) examination None N/A N/A Assembly Guide plates (cards) The specific relevant condition is wear that could lead to loss of control and alignment and impede control assembly insertion.

Control Rod Guide Tube All plants Enhanced visual (EVT-1) a. Bottom-mounted a. Confirmation of surface- a. For BMI column bodies, the Assembly examination instrumentation (BMI) breaking indications in two specific relevant condition for Lower flange welds column bodies or more CRGT lower flange the VT-3 examination is The specific relevant welds, combined with flux completely fractured column condition is a detectable b. Lower support thimble insertion/withdrawal bodies.

crack-like surface indication column bodies (cast, difficulty, shall require Not Applicable to visual (VT-3) examination b. For cast lower support PINGP 1&2), upper of BMI column bodies by column bodies (Not Applicable core plate and lower the completion of the next to PINGP 1&2), upper core support forging or refueling outage. plate and lower support casting forging/castings, the specific

b. Confirmation of surface- relevant condition is a breaking indications in two detectable crack-like surface or more CRGT lower flange indication.

welds shall require EVT-1 examination of cast lower support column bodies, upper core plate and lower support forging/castings within three fuel cycles following the initial observation.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 48 of 53 Attachment 2 MRP-227 -A Acceptance and Expansion Criteria Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Link(s) Acceptance Criteria (Note 1)

Core Barrel Assembly All plants Periodic enhanced visual a. Core barrel outlet a. The confirmed detection a and b. The specific relevant Upper core barrel flange weld (EVT-1) examination. nozzle welds and sizing of a surface- condition for the expansion core breaking indication with a barrel outlet nozzle weld and The specific relevant b. Lower support length greater than two lower support column body condition is a detectable column bodies (non inches in the upper core examination is a detectable crack-like surface indication. cast) barrel flange weld shall crack-like surface indication.

require that the EVT-1 examination be expanded to include the core barrel outlet nozzle welds by the completion of the next refueling outage.

b. If extensive cracking in the core barrel outlet nozzle welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 49 of 53 Attachment 2 MRP-227 -A Acceptance and Expansion Criteria Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Link(s) Acceptance Criteria (Note 1)

Core Barrel Assembly All plants Periodic enhanced visual a. Bottom-mounted a. The confirmed detection a. For BMI column bodies, the Lower core barrel flange weld (EVT-1) examination instrumentation (BMI) and sizing of a surface- specific relevant condition for (Note 2) column bodies breaking indication with a the VT-3 examination is The specific relevant length greater than two completely fractured column condition is a detectable inches in the lower core bodies.

crack-like surface indication. barrel flange weld,

b. Lower support b. For lower support forging or casting combined with flux thimble forging/castings, the specific insertion/withdrawal relevant condition is a difficulty, shall require detectable crack-like surface visual (VT-3) examination indication of BMI column bodies by the completion of the next refueling outage.

b. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the lower core barrel flange weld shall require EVT-1 examination of the lower support forging within three fuel cycles following the initial observation.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 50 of 53 Attachment 2 MRP-227-A Acceptance and Expansion Criteria Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Link(s) Acceptance Criteria (Note 1)

Core Barrel Assembly All plants Periodic enhanced visual Upper core barrel The confirmed detection The specific relevant condition Upper core barrel cylinder girth (EVT-1) examination cylinder axial welds and sizing of a surface- for the expansion upper core welds breaking indication with a barrel cylinder axial weld The specific relevant length greater than two examination is a detectable condition is a detectable inches in the upper core crack-like surface indication.

crack-like surface indication. barrel cylinder girth welds shall require that the EVT-1 examination be expanded to include the upper core barrel cylinder axial welds by the completion of the next refueling outage.

Core Barrel Assembly All plants Periodic enhanced visual Lower core barrel The confirmed detection The specific relevant condition Lower core barrel cylinder girth (EVT-1) examination cylinder axial welds and sizing of a surface- for the expansion lower core welds breaking indication with a barrel cylinder axial weld The specific relevant length greater than two examination is a detectable condition is a detectable inches in the lower core crack-like surface indication.

crack-like surface indication. barrel cylinder girth welds shall require that the EVT-1 examination be expanded to include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

REACTOR VESSEL INTEGRITY H44 H PROGRAM REV: 15 Page 51 of 53 Attachment 2 MRP-227 -A Acceptance and Expansion Criteria Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Link(s} Acceptance Criteria (Note 1)

Baffle-Former Assembly All plants with Visual (VT-3) examination. None NfA NfA Baffle-edge bolts baffle-edge bolts The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

Baffle-Former Assembly All plants Volumetric (UT) examination. a. Lower support a. Confirmation that more a and b. The examination Baffle-former bolts column bolts than 5% of the baffle- acceptance criteria for the UT of The examination acceptance former bolts actually the lower support column bolts criteria for the UT of the b. Barrel-former bolts examined on the four baffle and the barrel-former bolts shall baffle-former bolts shall be plates at the largest be established as part of the established as part of the distance from the core examination technical examination technical (presumed to be the lowest justification.

justification. dose locations) contain unacceptable indications shall require UT examination of the lower support column bolts within the next three fuel cycles.

b. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 52 of 53 Attachment 2 MRP-227-A Acceptance and Expansion Criteria Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Link(s) Acceptance Criteria (Note 1)

Baffle-Former Assembly All plants Visual (VT-3) examination. None N/A N/A Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high f1uence shroud plate joints, vertical displacement of shroud plates near high f1uence joints, and broken or damaged edge bolt locking systems along high f1 uence baffle plate joints.

Alignment and Interfacing All plants with 304 Direct physical measurement None N/A N/A Components stainless steel of spring height.

Internals hold down spring hold down springs The examination acceptance Not Applicable to criterion for this measurement PINGP 1&2 is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H44 H REACTOR VESSEL INTEGRITY PROGRAM REV: 15 Page 53 of 53 Attachment 2 MRP-227-A Acceptance and Expansion Criteria Examination Expansion Additional Examination Item Applicability Acceptance Criteria Expansion Criteria Link(s) Acceptance Criteria (Note 1)

Thermal Shield Assembly All plants with Visual (VT-3) examination. None N/A N/A Thermal shield fixtures thermal shields The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.

Notes to Acceptance and Expansion Criteria Table:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

2. The lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.