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- ELECTRIC POWER EPl21 RESEARCH INSTITUTE

MRP Materials Reliability Program MRP 2018-022 --------------

Date:

To:

From:

Subject:

August 31, 2018

MRP Research Integration Committee Members MRP Assessment TAC Members

David M. Czufin, PMMP Chair, TVA Brian Burgos, Program Manager, EPRI-MRP

Transmittal of MRP-191-SLR Screening, Ranking and Categorization Results and Interim Guidance in Support of Subsequent License Renewal at U.S. PWR Plants

An expert panel review supporting the Electric Power Research Institute (EPRI) Material Reliability Program's (MRP's} update of MRP-191 tor subsequent license renewal (SLR} was conducted in November 2017. The final publication of the revised report, MRP-191, Revision 2, is scheduled for CY2018. In order to support the application schedule of PWR utilities, prepublication versions of MRP-191, Revision 2 tables were developed based on preliminary/results from the expert panel review. These pre-publication tables are contained in the Enclosure to this letter.

This letter also transmits interim guidance that supplements MRP-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," [1] for subsequent license renewal

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(SLR}. This guidance for the Westinghouse and Combustion Engineering-designed reactor vessel internals is based on the current state of knowledge from the Electric Power Research Institute (EPRI} Materials Reliability Program (MRP} projects developing the guidance for SLR. (Note that similar guidance may be issued for Babcock and Wilcox-designed internals at a later date.} As of the publication of this interim guidance letter, MRP-175 [2] and MRP-211 [3] have been updated for SLR and published, providing the aging degradation screening criteria for developing MRP-227 for SLR. The update to MRP-191 [4], the "Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design," is currently underway. The expert panel review supporting a revision to MRP-191 has been completed and the proprietary meeting minutes published in the Westinghouse records management system. MRP-191, Revision 2 will be published by mid-2018. These preliminary results were used in developing this interim guidance.

The interim guidance provided in this letter is intended for use by utilities that are pursuing SLR prior to the publication of MRP-227 Revision 2 for SLR, which is anticipated for the end of 2020. Incorporation of these NEI 03-08 related Needed and Good Practice inspection requirements will NOT require implementation at any PWR plant prior to 12/31/2023. It is EPRl's understanding

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M RP 2018-022

that inspections in SLR PEO may begin no soonerthan calendar year 2029. Performance of these

SLR-related inspections do not need to be accelerated.

The guidance contained in the Enclosure is the current state of knowledge and is expected to

become applicable to the Westinghouse and Combustion Engineering fleet as appropriate by

publication of future guidance under the Materials Initiative as Interim Guidance or revision to

MRP-227 (e.g. Revision 1-A or Revision 2). It is noted that Section 2 of the Enclosure details the

applicability requirements for the interim guidance, and these requirements are the same as

those implemented for initial license renewal, using guidance previously promulgated in EPRI

letter MRP 2013-025. These applicability requirements are not new.

The Enclosure to this letter contains a summary of the technical basis for the recommended

changes and details of the recommended changes. Additional work remains during 2019-2020 to

complete the SLR-related updates associated with MRP-230 and MRP-232. The final

recommendations for.Primary, Expansion and Existing components requiring inspections during·

SLR PEO will be detailed in MRP-227-Revision 2, and may be different than those detailed herein.

INTERIM GUIDANCE FOR USING MRP-227-A TO SUPPORT SLR IN U.S. PWRs

The adjustments being made for SLR to MRP-227-A NEI 0~-08 Needed inspection tables for

Westinghouse-designed PWR plants include: (excerpted from Enclosure Tables 5-14 and 5-15)

Table 1: Expected New Entries for Table 4-3 for l\mP-227, Revision 2- Westinghouse Primary Components (NEI 03-08 Needed)

Primary Item Applicablllty Effect (Mechanism) Expansion Link Examination Method/ Frequency Examination Coverage

Alignment and All plants Cracking (SCC), Loss of None Visual (VT-3) examination no later than 2 All clevis insert

Interfacing material (Wear) refueling outages from the beginning of bolts and clevis

Components the first license renewal period. insert dowels

Clevis insert bolts Subsequent examinations on a ten-year

Clevis insert dowels interval.

Alignment and All plants with Loss of material (Wear) None Volumetric (UT) examination according Thermal sleeve wear

Interfacing thermal sleeves to the requirements and initial inspection surfaces according

Components timing ofTB-07-02 [29] and WCAP- to the requirements

Thermal sleeves 16911-P [40] of[29] [40] [41] Measurement of thermal sleeve guide funnels height according to the requirements and timing ofTB-07-02 [29] andPWROG-16003-P [41] Subsequent examinations based on calculated wear projections.

Radial Support Keys All plants Loss ofMaterial (wear) None Visual (VT-3) examination no later than 2 Wear surfaces on all

Radial support keys refueling outages from the beginning of radial support keys the first license renewal period. Subsequent examinations on a ten-year interval.

Alignment and All plants Loss ofMaterial (wear) None Visual (VT-3) examination no later than 2 Wear surfaces on all Interfacing refueling outages from the beginning of clevis inserts Components the first license renewal period.

Clevis bearing Stellite Subsequent examinations on a ten-year wear surface interval.

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M RP 2018-022

Table 2: Expected New :Entries for Table 4-9 for MRP-227, Revision 2 - Westinghouse Existing Components (NEI 03-08 Needed)

Item Applicability Effect (Mechanism) Reference Examination Method Examination Coverage

UCPandFuel All plants with Loss of Material (wear) ASME Code Section XI Visual {VT-3) All accessible surfaces at Alignment Pins malcomized fuel examination specified frequency

· Fuel alignment pins alignment pins on the UCP (See TB-16-4 [34])

LCPandFuel Allplants with Loss of Material (wear) ASME Code Section XI · Visual (VT-3) All accessible surfaces at Alignment Pins malcomized fuel examination specified frequency

Fuel alignment pins alignment pins on the LCP (See TB-16-4 [34])

The adjustments being made for SLR to MRP-227-A NEI 03-08 Needed inspection tables for Combustion Engineering (CE)-design PWR plants include: (excerpted from Enclosure Table 5-11)

Table 3: Expected Re,ised Entries for Table 4-2 for MRP-227, Re,ision 2 -CE Primary Components for Implementation in SLR (NEI 03-08 Needed) '

C11"ent Table Entries fMRP-227-A Table 4-2) Primary Item Applicability Effect (Medtanism) Expansion Link E!:amination l\Ietbod/ Enmination Conrage

Frequency

CoreShfoud Plant designs Distortion None Visual (VT-1) examination no If a gap exists. make three to Assembly with core (Void Swelling), as evidenced by later than 2 refueling outages five measurements of g;ip (Welded) shrouds

separation between the upper and from the beginning of the license opening from the core side at

Assembly assembled in

lower core shroud segments renm111 period. Subsequent the core shroud re-entrnnt two ,-ertical examinations on a ten-year corners. Then, evaluate the sections Aging :Management (IE) interval. swelling on a plant-specific

basis to determine frequency and method for additional requirements..

Core Support All plants Cracking (sec, IASCC) Lower cylinder axial welds Enhanced ,i.sual (EVT-1) 100"/o of the accessible Barrel

~Management (IE) examination no later than 2 su:daces of the lower cylinder

Assembly refueling outages from the welds.

Lower cylinder beginning of the license renew:il. period Subsequent examinations girth welds on a ten-yearinten111.

Rm'ised Table Entries Primary Item Applicability Effect (Mechanism) Expansion Link E!:amination l\Ietbod/ Enmination Coverage

Frequency

Con Shroud Plant designs Distortion None Visual (VT-1) examination no 100"/o of the horizontal seam Assembly with core (Void Swelling), as evidenced by later than 2 refueling outages between the upper and lower (Welded) shrouds

measurable separation between from the beginning of the fust core shroud segments.


the upper and lower core shroud license renewal period.

100"/o of the seam between two vertical Subsequent examinations on a sections

segmentsorbyshiftingofthe ten-year interval. the lower core shroud

segments relative to one another segment and the core support or the core support plate plate

Aging Management (IE)

Col'e Support All plants Cracking (SCC, IASCC) Aging Lower cylinder axial welds Enhanced visual (EVT-1) 100% of the accesstole Barrel Management (IE)

Fuel alignment plate (Plant examination no later than 2 smfices of.the lower cylinder

Assembly designs with core shrouds refueling outages from the welds. beginoing of the fust license Lower cylinder assembled in two vertical renewal period. Subsequent girth welds* sections only) examinations on a ten-year interval.

.. * Under MRP-227, Re11is1on 1 [2J, this component would be the Core Suppon Banel Assembly Middle Girth Weld (MGW) with expanstons to the mtddle axial weld (MAW) and lower axial ·weld (IA W). Per the responses to NRC RAis on MRP-227, Re,,i.sion 1 [15], the core support coltmmS could also become an expansion to the MGW. Once MRP-227, Revision 1 bas recei"-ed a safety evaluation "'ith acceptance of these changes, revisions to the ruuning of the Primary and Expansion componems fortheMRP-227-A ·1owe:r cylinder girth welds" pro,;idedin the approved versionofMRP-227, Revision 1 should be substituted here.

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M RP 2018-022

The majority of the recommended changes detailed in the Enclosure result from inspection

results and operating experience (OE) from domestic U.S. and international PWR units. These

changes are consistent with intent behind the NEI 03-08 materials initiative. Some of the

recommended changes are noted to apply during first license renewal period of extended

operation (PEO), and are included herein for completeness. These changes for first PEO will be

addressed in separate EPRI correspondence and/or through plant-specific implementation

actions at the affected plants.

Additional recommendations regarding asset management-related inspections for high economic

consequence components are also included in the Enclosure in Section 6. These are considered

as NEI 03-08 Good Practice items.

If you have additional questions or require further information, please contact Kyle Amberge

([email protected], 704-595-2039) or the undersigned.

Sincerely,

trA .. _. ·. :· . ~~ ~ . . . . ' . QJ. g-··,.

David M. Czufin PMMP Chair

TVA

CC: PMMP Members

Brian Burgos MRP Program Manager

EPRI

ENCLOSURE: Interim Guidance for the Pressurized Water Reactor Internals Inspection and

Evaluation Guidelines, MRP-227-A, for Subsequent License Renewal-Westinghouse and

Combustion Engineering-Designed Reactor Vessel Internals

References:

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation

Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

2. Materials Reliability Program:PWR Internals Material Aging Degradation Mechanism

Screening and Threshold Values (MRP-175, Revision 1). EPRI, Palo Alto, CA: 2017. 3002010268.

3. Materials Reliability Program: PWR Internals Age-Related Material Properties, Degradation

Mechanism, Models, and Basis Data-State of Knowledge (MRP-211, Revision 1). EPRI, Palo

Alto, CA: 2017. 3002010270.

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MRP 2018-022

4. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191, Revision 1). EPRI, Palo Alto, CA: 2016. 3002007960

5. Westinghouse Electric Company Letter, LTR-AMLR-17-28, Revision 1, "MRP-191 Revision 2 Expert Panel Meeting Minutes," June 15, 2018 (Westinghouse Proprietary).

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Enclosure to MRP 2018-022 ~~1211 ELECTRIC POWER ~·- RESEARCH INSTITUTE

Interim Guidance for the Pressurized Water Reactor Internals Insp~ction and Evaluation Guidelines, MRP-227-A,

for Subsequent License Renewal-Westinghouse and Combustion Engineering-Designed Reactor Vessel Internals

. Prepared for EPRI by Joshua K. McKinley Bradley T. CarpenteF

Bryan M. Wilson John F. Kielb

Patrick M. Minogue AmyE. Freed

John L. McFadden Westinghouse Electric Company

(

Ref: Westinghouse Electric Co. letter LJR-AMLR-18-4, Rev. I, dated 7/9/2018

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Enclosure to MRP 2018-022

TABLE OF CONTENTS

# Section Page

1 Introduction and Purpose 1

2 Applicability Requirements for the Interim Guidance 5

3 Key Factors Impacting the Development ofMRP-227 for SLR 7

4 Results of the MRP-191 SLR Expert Panel Review 11

5 Expected impacts to MRP-227 for SLR 25

5.1 Impacts on CE Inspection Requirements 31

5.2 Impacts on Westinghouse Inspection Requirements 43

6 Asset Management Recommendations for License Renewal 58

7 Conclusions 67

8 References 68

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Westinghouse Non-Proprietary Class 3 Enclosure to MRP 2018-022

Page 1 of70 LTR-AMLR-18-4, Rev. 1

July 9, 2018

Interim Guidance for the Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227, for Subsequent License RenewalWestinghouse and Combustion Engineering-Designed Reactor Vessel

Internals

1 Introduction and Purpose

The current generation of pressurized water reactor (PWR) plants is approaching the end of their respective licensing periods, and multiple plants have already entered their first period of extended operation (PEO). The nuclear power industry in the United States has developed inspection and evaluation (i&E) guidelines for managing aging degradation in reactor vessel internals for the first PEO. The Nuclear Regulatory Commission (NRC) approved version of these guidelines is published in MRP-227-A [1], while the most current version (unapproved) is published in MRP-227, Revision 1 [2].

Several utilities have declared their intent to pursue subsequent license renewal (SLR) to extend plant licenses beyond the first PEO, which is beyond the scope ofMRP-227-A and MRP-227, Revision 1. To update MRP-227 for SLR, the technical basis documents supporting reactor internals aging management strategy development must also be updated. This task is already underway, with revisions to MRP-175 [3] and MRP-211 [4] being published in mid-2017. These two documents updated the materials aging degradation database with the most recent developments and testing and revised the screening thresholds for some of those degradation mechanisms. Work is currently underway to update the next basis document in the series, MRP-191: "Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design" [5]. An expert panel review was held in November 2017 to develop the screening, categorization, and ranking tables for the revised

. MRP-191 and was supplemented by a follow up expert panel review on May 29, 2018 to address recent operating experience [6]. The MRP-191 revision is scheduled for publication in mid-2018. Figure 1-1 shows the process currently being implemented to develop MRP-227 for SLR. The development of MRP-191, Revision 2 is captured by the top portion of the flow chart labeled "Categorization."

The publication ofMRP-227 for SLR is not expected until 2020, which does not support the utility owners of the early plants for SLR in developing their SLR applications, Even the schedule for publication of the SLR revision to MRP-191 (Revision 2) does not support the lead SLR plants. The

· utility owners of the lead plants are actively preparing their SLR applications now and may even submit those applications prior to MRP-191, Revision 2 publication.

These early applicants could develop their own plant-specific approaches to managing the aging of the reactor vessel internals. A range of approaches could be used by these utilities. For example, the MRP-227-A [I] or MRP-227, Revision 1 guidance [2] could be used with some reasonable changes based on knowle.dge from MRP-175, Revision 1 [3] and plant-specific analyses. A utility could even develop its own approach from scratch. However, the best practice has historically been for the industry to consistently apply the same guidance at each plant site, if possible. Since significant portions of the technical basis for MRP-227 for SLR have already been developed through the publication ofMRP-175, Revision I [3] and MRP-211, Revision I [4], the completion of the expert panel review for MRP-191,

© 2018 Westinghouse Electric Company LLC All Rights Reserved



July 9, 2018

Revision 2 [6], and the earlier publication of the existing I&E guidelines [1] [2], an initial version of the

I&E guidance can be created based on current knowledge.

The purpose of this interim guidance letter is to provide the lead and early SLR plant owners with a

reactor vessel internals aging management template based on the technical foundation work that has been

completed to date. This will support the development of SLR reactor vessel internals aging management

programs for those plants to meet the requirements of NUREG-2191 [7] and NUREG-2192 [8]. This also

helps those plant owners to achieve consistency and technical rigor in their SLR applications created prior

to the final publication of MRP-227 for SLR.

As with the guidance provided in MRP-227, this interim guidance will be required under the NEI 03-08

framework [9]. The aging management recommendations based on safety will be classified as NEI 03-08

"Needed" requirements, while the asset management recommendations based on economic risk will be

classified as NEI 03-08 "Good Practice" requirements. The descriptions of these two NEI 03-08

categories are provided here for corivenience:

Needed - to be implemented wherever possible, but alternative approaches are

acceptable. Criteria that qualify an element of a work product as "Needed" include:

• Element substantively affects the ability of structures, systems or components to

reliably perform their economic function.

• Element would be moderately risk significant as determined by the responsible IP

[issue program] if not implemented.

• Element addresses a material degradation mechanism that has significant financial

impact on the entire industry, especially where failure at one plant could affect many

other plants.

• A consensus of the responsible materials IP believes the element should be

designated as "Needed".

Good Practice - implementation is expected to provide significant .operational and

reliability benefits, but the extent of use is at the discretion of the individual utility.

Specific elements of a work product that may be assigned this criterion include:

• Element reflects an industry standard of performance or represents a consensus

opinion of the responsible materials IP.

• A consensus of the responsible materials IP believes the element should be

designated as "Good Practice".

·This interim guidance is based primarily on the following references available at the time of its

publication:

• MRP-227-A: NRC-approved I&E Guidelines [1]

• MRP-227, Revision 1: I&E Guidelines [2]

• MRP-175, Revision 1: Aging degradation mechanism screening thresholds [3]

• MRP-211, Revision 1: Aging degradation mechanism basis data [4]



July 9, 2018

• MRP-191, Revision 1 : Screening, categorization, and ranking of internals components [ 5] • L TR-AMLR-17-28, Revision 1 : SLR Expert panel review results for MRP-191, Revision 2 [ 6] • MRP 2017-009: Baffle-former bolt interim guidance [10] • MRP 2018-002: Baffle-former bolt expansion criteria for large clusters [11] • MRP 2018-007: Guide card wear interim guidance [12] [13] • MRP 2017-027: MRP-227, Revision 1 request for additional information (RAI) responses [14] • MRP 2018-003: MRP-227, Revision 1 RAI responses [15] • MRP-2018-011: Responses to supplemental NRC questions on MRP-227, Revision 1 [16] • Various operating experience references

Per NUREG-219Y [7] and NUREG-2192 [8], MRP-227-A should be the initial reference basis for developing .a reactor vessel internals aging management program for SLR. This reference basis must be supplemented with a gap analysis to determine the changes required to provide reasonable assurance that the aging degradation effects will be managed during SLR. · The documents listed above, particularly the; revisions of MRP-227 and the expert panel review results, are the key ipput references for this gap analysis as of the publication of this interim guidance letter.

If applying this interim guidance, it must be recognized that the conclusions and recommendations provided here are based on the currently available· knowledge. When the final MRP-227 for SLR is issued, it will be based on the current information available at that time and could be different from the information provided here. This risk has been minimized as much as possible by incorporating all of the relevant current knowledge from the existing documents and expert panel reviews listed above. However, there are several sources of non-negligible risk that could cause changes between this interim guidance letter and the final publication ofMRP-227 for SLR around 2020. These include the current NRC review ofMRP-227, Revision 1 [2], potential operating experience events for reactor internals components, and additional information gathered during the remaining steps in developing the l&E guidelines. This last source of risk is demonstrated by the flow chart in figure 1-1, which shows that two· additional steps remain after the development ofMRP-191, Revision 2: the "Analysis" and "Aging Management Strategy Development" steps .

. Therefore, even though this interim guidance provides requirements under the NEI 03-08 framework, utility owners tµat apply this interim guidance should continue to follow the development .of the final

. industry I&E guideJines for SLR, .because additional or revised requirements may be included. Once the final guidelines are published, the aging management programs implemented based on this interim guidance may require revision to reconcile the differences.

Similar to MRP~227-A/Revision 1 [1] [2], this interim guidance prescribes programs and activities that will assure the long-term safe and reliable operation of PWR internals as they age. One difference for this revision of the l&E guidelines is that the impacts of economic risks relevant to asset management have been separated from the impacts of safety risks. To make this separation clear, the aging management recommendations re:levant to safety are provided in one section of this report, while the recommendations relevant only to·asset management are provided in a separate section with a separate set of implementation tables.

Enclosure to MRP 20 18-022 Westinghouse Non-Proprietary Class 3

Page 4 of70 LTR-AMLR-18-4, Rev. I

July 9, 2018

This guidance is applicable to the reactor internals structural components: items such as fuel assemblies or reactivity control assemblies are not addressed. This guidance does NOT supersede or modify any plant-specific commitments without specific approval to do so by the regulatory body (commitments related to IO CFR 54, ASME Section XI, etc.), and it does NOT modify any previous NEI 03 -08 guidance unless specifically addressed by this interim guidance or future revisions of MRP-227. The interim guidance also does NOT apply to new plants beginning construction after calendar year 2007.

Component List

Below Screening

No Credible Damage Issue

Screening Criteria

Categorization

Initial Screening

Above Screening

Likelihood & Severity Analysis

Moderate Cat. A

No Adverse Effects

No Add it ional Measures

.--------............ High

Category B

Resolved by Analysis

Ag ing Management

Expansion

Category C

Anal Engineering Evaluation

and Safety Assessment

Aging Management

Strategy ,____.._D_,evelopment

Existing Programs New Requirements

Aging Management Program l&E Guidelines

Figure 1-1: Process fo r Categor ization of PWR Internals Components for MRP-227

Westinghouse Non-Proprietary Class 3

2 Applicability Requirements for the Interim Guidance



July 9, 2018

This interim guidance is not intended to cover every possible plant-specific license renewal or power uprate commitment. Plant-specific commitments remain the responsibility of the utility owner. Similar to MRP-227-A/Revision 1 [1] [2], the technical basis documents for MRP-227 for SLR are being developed based on a broad set of assumptions about plant operation, which encompass the range of current plant conditions for the domestic fleet of PWRs. These applicability requirements are generally the same as those provided in the earlier versions ofMRP-227, with the addition of the supplemental applicability requirements provided in MRP 2013-025 [17] to respond to MRP-227-A applicant/licensee action items imposed by the NRC.

The engineering analyses and evaluations that support this interim guidance were based on representative configurations and operational histories, which were generally conservative, but not necessarily bounding in every parameter. Users of this interim guidance are expected to confirm with reasonable assurance that each reactor managed with the guidance satisfies the following assumptions for the power plant: ·

• Has operated for 30 years or less with high-leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management · strategy for the remaining years of operation-the limitations defining low leakage operation for SLR considerations are the same as those used for the first PEO:

o Combustion Engineering (CE) [17] [18] • Heat generation rate figure of merit: F ~ 68 Watts/cm3

• Average core power density< 110 Watts/cm3

• · Active fuel to fuel alignment plate distance> 12.4 inches o Westinghouse [17] [19]

• Heat generation rate figure of merit: F ~ 68 Watts/cm3

• Average core power density< 124 Watts/cm3

• Active fuel to fuel alignment plate distance> 12.2 inches • Has operated for the majority of its lifetime as a base-loaded unit and is currently operating as a

base loaded power plant (unit operates_ at fixed thermal power levels and does not usually vary power on a calendar or load demand schedule)

• Has not implemented design changes beyond those identified in general industry guidance or recommended by the original vendors

• Has confirmed that the components and material class of each functional component are as listed in the latest revision ofMRP-191, as applicable to the individual plant design

o Note that this includes confirmation that the components fabricated from austenitic stainless steel which are not welds or bolting components do not contain 20 percent or greater cold work [ 17]

o The final safety evaluation on Applicant/Licensee Action Items 1 and 7 from the safety evaluation on MRP-227-A supports resolutio~ of this applicability requirement [20]

These assumptions are a conservative representation of U.S. PWR operating plants. Note that any plants which have implemented flexible operation for extended periods either under the original license or a PEO are not covered by this interim guidance. This interim guidance for SLR can still be used at these



July 9, 2018

plants, but the utility owners of those plants must justify application of the guidance through

consideration of the impacts of flexible operation.

The recommendations are applicable to all U.S. PWR operating plants as of November 2017 for the three

designs identified (with the flexible operations exception noted above). These guidelines are also

considered applicable to plants that have replaced components or component assemblies as noted in the

1applicability guidelines above; however, the impact c,f such design changes may require evaluation for

how factors such as differing component age will impact the I&E guidance.

If major plant-specific differences from the inputs to the failure modes, effects, and criticality analysis

(FMECA) process described here or in MRP-191 [5] are identified, then plant owners must determine and

document the impact, if any, on the aging management strategy described herein.

Plant modifications to PWR internals ( e.g., physical changes) made after calendar year 2017 should be

reviewed to assess impacts on strategies_contained in these guidelines.

MRP-227-A/Revision 1 [1] [2] identified certain CE and Westinghouse PWR internals components

subject to inspection under existing programs as requiring further plant-specific evaluation to verify the

acceptability of the existing programs or to identify changes to those existing programs which should be

implemented to manage the aging of these components for the period of extended operation. These were

generally components managed by existing programs other than ASME Section XI examinations. If the

existing programs are not acceptable, it is necessary to identify and implement changes to the programs to

manage aging of applicable components over the period of extended operation. It is the owner's

responsibility to perform a plant-specific evaluation to verify the acceptability of the existing programs,

or to identify changes to the program or component replacement strategies and initiatives that should be

implemented to manage the aging of these components for the period of extended operation.



July 9, 2018

3 Key Factors Impacting the Development of MRP-227 for SLR

MRP-227-A/Revision 1 [I] .[2] were developed for the first PEO in order to extend the original licenses of the Westinghouse and CE PWRs from 40 years to 60 years. The basis documents ofMRP-227, such as MRP-191, Revision 1 [SJ, also assumed this 60-year operating timeframe. Increasing plant operation tfme. to 80 calendar years could impact the conclusions documented in the I&E guidelines because the aging degradation mechanisms have had additional time to evolve in the reactor vessel internals components and because the neutron irradiation dose has continued to increase in the components. The conclusi.ons could also be impacted by developments in the understanding of the aging degradation mechanisms since the original publication of MRP-175, Revision O [21] or by operating experience events that have occurred since the original publication ofMRP-227 and its basis documents.

, The eight aging degradation mechanisms that can have an impact on the reactor vessel internals were investigated in detail in MRP-175 [3] [21]. The following mechanisms were included:

• Stress corrosion cracking (SCC) • Irradiation-assisted stress corrosion cracking (IASCC) • Wear • Fatigue • Irradiation embrittlement (IE) • Thermal embrittlement (TE) . • Irradiation-induced stress relaxation and creep (ISR/IC) • Void swelling (VS)

SCC, IASCC, wear, fatigue, and thermal embrittlement are all time-dependent, increasing with additional time spent exposed to the environment. IASCC also has an i~pact from radiation dose accumulation, since it does not occur at zero or very low dose but does occur at higher doses. IE, ISR/IC, and VS. are all dose-dependent, increasing with increased radiation exposure. Some of these effects, particularly IE and TE, typically reach a saturation level after enough dose exposure or time. Increasing the assumed operational time of a PWR from 60 years to 80 years increases both time of exposure and the accumulated radiation dose. These were both evaluated in preparing the inputs to the MRP-191 SLR expert panel review and as part of the review itself [6]..

The first step of this was the publication ofMRP-175, Revision 1 [3] where the screening thresholds for each of the eight aging degradation mechanisms were updated with consideration of recent testing and operating experience and the extension of the plant life to 80 years. The previous MRP-17 5 screening thresholds [21] were quite conservative and most did not include an impact of the length of time the piece was exposed. For example, SCC was based on the applied stress and the condition of the material and not the length of time that the component was exposed to the applied stress. Thus, only two degradation mechanisms required updates to thefr screening thresholds: IASCC and fatigue. The threshold for IASCC required an update because of new laboratory data published sirice MRP-175, Revision 0, which resulted in a new trend curve for IASCC. The threshold for fatigue required an update due to the combination of additional time to operate to 80 years (i.e., additional fatigue cycles) and recent developments in understanding the environmental effects on fatigue, which together resulted in a lower cumulative usage factor (CUF) for the fatigue threshold. The new trend curve for IASCC in MRP-175,



July 9, 2018

Revision 1 did not impact the screening because screening criteria used for MRP-191, Revision O ·and

Revision 1 (see Section 5.1.2 and Figure 5-1) [5] [23] were more conservative than the new IASCC curve.

The SLR expert panel review continued to use the same conservative IASCC screening criteria as

MRP-191, Revision O and Revision 1. The lower CUF threshold for the fatigue screening did result in a

significantly larger number of screened in components; however, the expert panel determined that the

lower screening criteria did not result in promotion of new lead components for fatigue: ' .

The second step was to address the increases to neutron irradiation dose at 80 years. Calculations were ·

performed for representative CE [18] and Westinghouse [19] designed plants. Neither of these

calculations can be. called "bounding" due to the fact that dose calculations were not performed for every

operating plant or every possible future operation case. However, steps were taken to add an appropriate

level of conservatism to each of the calculated dose maps in order to address the different designs and the

potential variations. Additionally, the applicability criteria for core operation originally provided in MRP

2013-025 [17] and repeated in Section 2 of this report were maintained in the 80-year calculations. The

following steps were taken to obtain representative dose projections with a reasonable amount of added

conservatism:

• CE plants: Dose projections were generated using a model developed for one specific plant at 72

effective full-power years (EFPY). The inputs for that plant were compared to information

available for several other CE plants and confirmed to be appropriately conservative. To account

for variations in axial and radial power shapes, two different dose projections were generated:

a) A flat axial power shape that produced conservative results above and below the active

fuel b) A best-estimate power shape with 30% margin added that was more limiting in the radial

direction

A composite dose map was generated using the maximum value of a) and b) for each mesh cell in

the neutron transport calculation.

• Westinghouse Plants: Dose projections were generated using an available model developed for

one 3-loop plant and a second available model developed for one 4-loop plant at 72 EFPY. To

account for variations in axial and radial power shapes, two different dose projections were

generated for each representative plant:

a) A flat axial power shape that produced conservative dose projection results above and

below the active fuel

b) A best-estimate and realistic axial power shape that produced dose projection results that

were more limiting in the radial direction

The dose projections from both a) and b) were adjusted by conservative margin factors based on

comparisons made between the inputs for the base model plants and other similar plants. In the

case of the 3-loop plant, the available model was not for the plant with the highest inputs, so a

multiplier of 1.3 was applied to the flat axial distribution and a multiplier of 1.6 was applied to

the best-estimate distribution. In the case of the 4-loop plant, the plant modeled had more

conservative inputs, so the flat axial distribution results were used without a multiplier and the

best-estimate distribution was multiplied by 1.3.

Westinghouse Non-Proprietary Class 3 ~nclosure to MRP 2018-022


July 9, 2018

As with the CE plants, a composite dose map was generated using the maximum value of a) and b) for each mesh cell in the neutron transport calculation.

The revised MRP-175, Revision 1 [3] screening thresholds1and the 80-year neutron dose proje.ctions were the key inputs to updating the screening inputs (MRP-191, Tables A-1 and A-2 [5]) and screening results (MRP-191, Tables 5-1 and 5-2 [5]) for-the MRP-191 SLR expert panel [22]. These tables were provided to the expert panel as an input for developing the categorization and ranking of the reactor vessel internals components for SLR.

The original I&E guidelines incorporated the experts' knowledge of applicable operating experience up to I

'

th~t time. Operating experience with reactor vessel internals components and materials has generally been positive, a trend which has continued since the original MRP-191, Revision O [23] expert panel

. review. The SLR expert panel considered several relevant developments in operating experience for its review of the components: ·

• Guide plates/cards: wear of the guide plates/cards in Westinghouse-designe~ plants was a known issue during the original MRP-191 expert panel review, with the guide cards being assigned to Category C for wear [23] and then becoming a primary inspection item in MRP-227-A/Revision 1 [l] [2]. Wear has been observed at multiple other plants [24] [25] and unexpected results for specific design types, like the accelerated wear due to ion nitride rod cluster control assemblies [26], have also been observed. Interim guidance for the most recent operating experience was just recently published in letter OG-18-76 [13].

• - Baffle-former bolts: IASCC degradation of the baffle-former bolts in Westinghouse-desigried plants was also a known issue during the original MRP-191 expert panel review, with the bolts being assigned to Category C in particular for IASCC and fatigue and then becoming a primary inspection item in MRP-227-A/Revision 1. Cracking degradation of the baffle-former bolts has been observed at multiple plants. Since the original expert panel review, a more severe iteration of baffle-former bolt degradation has been observed at multiple plants, where large clusters of adjacent bolts fail [27] [28]. This is thought to coincide with an acceleration of the rate ofbaffleformer bolt failures. The operating experience with baffle-former bolts was considered relevant for not only the baffle-former bolts but also for other similar bolts, like the lower support column bolts and barrel-former bolts, which have similar designs and have accumulated significant radiation dose.

• Reactor vessel head adapter thermal sleeve: wear has been observed at several Westinghousedesigned plants in the reactor vessel head adapter thermal sleeves.· Unacceptable levels of wear have been observed in several cases and in 2014, a thermal sleeve fell from·the reactor vessel closure head because the upper flange on the sleeve had worn through and no longer held it in position. This wear mechanism potentially affects multiple plants [29] [30]. Recent operating experience at an Electricite de France (EdF) plant has shown' the potential for thermal sleeve degradation to result in stuck control rods, causing a possible safety hazard [31] [32]. In response, EdF is reviewing the remote visual inspections of the affected areas in all 900 MW and 1300 MW units of the EdF fleet.

(



July 9, 2018

• Clevis insert bolts: SCC degradation of the bolts that hold the clevis inserts in place in

Westinghouse-designed plants has been observed [33]. This has been attributed to the susceptible

Alloy X-750 material of the bolts and primary water SCC.

• Clevis inserts and radial support keys: wear of the alignment surfaces provided by the clevis

insert and the radial support keys in Westinghouse-designed plants has been observed in some

plants. This wear was listed as the focus for the existing inspection included in MRP-227-

A/Revision 1 and in the technical bulletin documenting the clevis insert bolt degradation [33].

This type of wear was also considered possible for other similar items in the Westinghouse and

CE-designed reactor vessel internals.

• Malcomized fuel alignment pins: surface degradation of fuel alignment pins that were fabricated

with a malcomized surface hardening treatment has been observed [34]. This appears as a flaking

off of the surface material. This could lead to accelerated wear and loss of the material on the

surface.

• Brackets, clamps, terminal blocks, and conduit straps: during the SLR expert panel review, it was

noted that the clamps covered by this component have experienced some aging degradation

failures [6]. It was thought that this was likely due to SCC of cold worked regions created during

installation. • Recent operating experience: during the late 2017 and early 2018 outages (after the November

2017 expert panel meeting was held), there were three operating experience events that could

have an impact on the recommendations of this letter. Due to how recently these occurred,

references and extensive details are not available at the publication of this interim guidance. The

May 2018 expert panel was held to re-evaluate the affected components in light of this operating

experience.

r

o · Interference of a degraded thermal sleeve with the insertion of a control rod: This

increased the safety risk of thermal sleeve degradation.

o Degradation of a core .shroud tie rod at a CE-design plant: a core shroud tie rod was

found to have dropped from its design position.

o Crack-iike indications observed in a CE-design plant core support barrel: the indications

were observed through an MRP-227-A examination and supplemental examinations

performed after the initial observations. Note that this interim guidance does not create

revised requirements for CE- or Westinghouse-designed core barrels, since such guidance

would be premature. The industry is currently evaluating the need for further guidance

on core barrel aging management.

The updates to the degradation mechanism screening inputs and screening results and the updated

operating experience knowledge provided to the StR expert panel each had some impact on the results of

the expert panel review and other supporting evaluations. These items were factored into the review by

the expert panel and supporting evaluations and form the foundation for developing the interim guidance.


4 Results of the MRP-191 SLR Expert Panel Review



July 9, 2018

The MRP-191 SLR expert panel review was conducted on November 7-9, 2017. A follow up expert panel review was conducted on May 29, 2018 to address the most recent operating experience for reactor vessel internals. The results of both.of these expert panel reviews have been documented in the meeting minutes [6] and will eventually be used as input to Revision 2 ofMRP-19L As noted in Section 3, updated screening inputs and screening results [22] were provided to the expert panel for input to the discussions. The expert panel review produced several of the key MRP-191 tables:

• Table 6-5: FMECA Results for Westinghouse Reactor Internals • Table 6-6: FMECA Results for CE Reactor Internals • Table 7-2: Categorization of Westinghouse Reactor Internals Components • Table 7-3: Categorization of CE Reactor Internals Components

The review was governed by the expert panel review process and definitions documented in MRP-191 Sections 6 and 7 [5], with a few changes intended to improve the flexibility and usefulness of the results. For the SLR expert panel review, most of this process was implemented in the same way as for the first PEO. Two significant changes were made:

• The consequence of degradation was separated into two separate categories: safety consequence and economic ·consequence. These were combined for the previous FMECA, which caused difficulty when attempting to discern whether a high consequence category referred to safety risks or economic risks. Because of this separation, the FMECA group and the risk categories were also separated into safety and economic versions.

o Safety consequence FMECA category descriptions are shown in Table 4-1 o Economic consequence FMECA category descriptions are shown in Table 4-2 o Safety consequence risk category descriptions are shown in Table 4-3 o Economic consequence risk category descriptions are shown in Table 4-4

• Table 6-4 ofMRP-191 [5] was modified to be more conservative in its results for high likelihood events. The original version of the table is shown in Table 4-5. The revised table for the SLR expert panel is shown in Table 4-6. This revision served to put a higher emphasis on the most likely events to occur or those events that have already occurred and caused the FMECA to produce more conservative results.

These two changes were intended to improve the potential for the MRP-227 SLR I&E guidelines to act as an effective reactor vessel internals aging management tool. The separation of the consequence of degradation into two sub-categories also has the benefit of supporting effective use of the results for purposes focused individually on safety risk or on economic risk. The separation will be used in the following sections of this interim guidance to provide component aging management strategies recommended for addition to the MRP-227, Section 4 inspection tables as well as asset management strategies intended to reduce economic risk for utilities.


Page 12 of70 LTR-AMLR-18:4, Rev. 1

July 9, 2018

The component has no screened-in degradation mechanism. No need to assess core damage

probability

Expert panel believes there is no credible means for component failure(s) to cause core

damage

Expert panel believes {he potential exists for core damage as a result of component ( or

multiple) failure(s) but that the ability to shut down the reactor in a controlled manner

Expert panel believes that some core damage could possibly result from failure of the

component( s)

Cost that can be generally handled within the existing plant outage budget and resources

(<$SM)

Cost that exceeds the normal plant outage budget and resources t>$5M)

Cost that potentially affects the utility's overall enterprise/financial health (>$20M)


I Enclosure to MRP 2018-022


July 9, 2018

Table 4-3: Safety Consequence Risk Cate2ory Descriptions Category Description

I

Category A Those component items for which aging effects are below the screening criteria. Aging degradation significance is minimal. The initial set of Category A components consists of items for which all degradation mechanisms are screened out. These components are identified as "None" in the appropriate columns of the screening and categorization tables.

-In addition, the FMECA results can identify additional components for which age-related degradation mechanisms have minimal likelihood to cause failure. These components are also assigned to Category A. This action essentially screens these components out of further consideration for future steps in developing MRP-227 for SLR. Additional components may ultimately be categorized as Category A as discussed in the Category B

•. definition.

Category C Those "lead" component items for which aging effects are above screening levels. Aging degradation significance is high or moderate. Enhanced/augmented inspections and/or surveillance sampling typically may be warranted to assess aging effects and verify component item safety functionality.

These components, for which aging effects are above the threshold values of the screening criteria, are assessed to have moderate to high likelihood of occurrence, and have the potential for significant damage. Moreover, they have not been demonstrated, analytically or by experiment, to be sufficiently damage-tolerant to remain functional relative to the aging degradation mechanism(s) identified.

CategoryB Category B items are defined as those component items that also are above screening levels

' but are no! "lead" component items. Aging degradation significance is moderate. Category B component items may require additional evaluations to be shown tolerant of the aging effects with no loss offunctionality (i.e., damage tolerant).

Non-Category A components that are judged to have moderate susceptibility and potentially significant consequences, such that the effects on function cannot easily be dispositioned by screening, and yet are not considered Category C components, are assigned to Category B. Some of the components included in the Category B list may have been screened in for susceptibility to one or more degradation mechanisms, but the likelihood of occurrence and the implied safety risk were assessed by qualitative expert assessment to be low to moderate. If it is further concluded that the" existing 10-year in-service inspection or other in-place aging management plans are sufficient to preclude a safety concern, such components can be reassigned as Category A.



July 9, 2018

Table 4-4: Economic Consequence Risk Category Descriptions

Category Description

Category A Those component items for which aging effects are below the screening criteria. Aging

degradation significance is minimal. The initial set of Category A components consists of

items for which all degradation mechanisms are screened out. These components are

identified as "None" in the appropriate columns of the screening and categorization tables.

In addition, the FMECA results can identify additional components for which age-related

degradation mechanisms have minimal likelihood to cause failure. These components are

also assigned to Category A. This action essentially screens these components out of

further consideration for future steps in developing MRP-227 for SLR. Additional

components may ultimately be categorized as Category A as discussed in the Category B

definition.

Category C Those "lead" component items for which aging effects are above screening levels. Aging

degradation significance is high or moderate. Enhanced/augmented inspections and/or

surveillance sampling typically may be warranted to assess aging effects, verify component

item functionality, and identify extent ofrepairs that may be required (including likely cost

and outage duration impacts).

These componeI).ts, for which aging effects are above the threshold values of the screening

criteria, are assessed to have moderate to high likelihood of occurrence, and have the

potential for significant economic or reliability consequences. Moreover, they have not

been demonstrated, analytically or by experiment, to be sufficiently dam1;1ge-tolerant to

remain functional relative to the aging degradation mechanism(s) identified.

CategoryB Category B items are defined as those component items that also are above screening levels

but are not "lead" component items. Aging degradation significance is moderate. Category

B component items may require additional evaluations to be shown tolerant of the aging

effects with no loss offunctionality (i.e., damage tolerant).

Non-Category A components that are judged to have moderate susceptibilit)'. and

potentially significant economic consequences, such that the effects on function cannot

easily be dispositioned by screening, and yet are not considered Category C components,

are assigned to Category B. Some of the components included in the Category B list may

. have been screened in for susceptibility to one or more degradation mechanisms, but the

likelihood of occurrence and the implied economic risk were assessed by qualitative expert

assessment to be low to moderate. If it is further concluded that the existing 10-year in-

service inspection or other in-place aging management plans are sufficient to preclude a

reliability or functional concern, such components can be reassigned as Category A.

'\ I



July 9, 2018

Table 4-5: MRP-191, Revisions O and 1 [23] [5] Reactor Internals FMECA (Significance) Groups

Failure Consequence (Damage Likelihood) Likelihood Low Medium High

High 2 3 3 Medium 1 2 3

Low 1 1 2 None 0 0 0

Table 4-6: MRP-191 SLR Expert Panel Reactor Internals FMECA (Significance) Groups

Failure Consequence (Damage Likelihood) Likelihood . Low Medium High

High 3 3 3. Medium 2 2 3

Low 1 1 2 None 0 0 0

The outputs generated by the expert panel review were developed based on a specific set of inputs. These . inputs include the list of components in scope, the materials of fabrication for those components, the calculated neutron fluence and dose, the stress and operating conditions assumed, and the known operating experience and plant modifications. These inputs resulted in the applicability requirements provided in Section i In order to use the results of the1SLR expert panel review and any downstream outputs, a utility owner must show that the applicability requirements of Section 2 are met.

Similar to the expert panel outputs documented in MRP-191, Revision 1 [5], the SLR expert panel review resulted in a number of components being assigned to the highest risk category. Some of these were for both the safety and economic risk categories but more were for economic consequences alone. The components that were assigned to Category C by the SLR expert panel are provided in Table :4-7 for the CE components and in Table 4-8 for the Westinghouse components. The SLR Category C co"mponents are compared to the MRP-191, Revision 1 Category C components in Table 4-9 and Table 4-10 for CE and Westinghouse components, respectively.

Significantly more components were assigned to Category C for the Westinghouse design than the CE design. This is primarily motivated by the larger amount of degradation observed to date in the operating experience for PWRs of Westinghouse design.

There are a total of ten components in Category C for the CE design (Table 4-7) and 2.1 in Category C for Westinghouse design (Table 4-8). The ratio between the two designs is similar to the results ofMRP-191, Revision 1, where there were six components in Category C for CE design (Table 4-9) and eleven in Category C for Westinghouse design (Table 4-10).



July 9, 2018

Comparing the new results from the SLR expert panel review to the previous results from MRP-191, Revision 1 shows that five components were added to the group of CE Category C items (Table 4-9):

• Fuel alignment plate • Upper cylinder (includes UFW, UGW, and UAW) • Core stabilizing lug shim bolts • Core shroud tie rods • Core shroud tie rod nuts

These five components were only added as economic Category C items. Three'Category C items from

MRP-191, Revision 1 were reduced to lower risk categories:

• Lower support structure fuel alignment pins

• Core shroud plates • In-core instrumentation (ICI) thimble tubes-lower

The notes with Table 4-9 provide details on why individual components were increased or decreased in

risk category. The revised screening criteria, recent operating experience, and increases in time and

neutron dose influenced the increases in risk category. The decreases in risk category were related to modifications that have already addressed the aging degradation issue, such as with the ICI thimble tubes

lower and fmprovements in the understanding of the component and degradation meshanisms, such as the

core shroud plates due to the reduction in the expected VS [35].

Comparing the new results from the SLR expert panel review to the previous results from MRP-191,

Revision 1 shows that ten components were added to the group of Westinghouse Category C items (Table 4-10):

• Economic Risk Category C only: o Upper core plate (UCP) o Brackets, clamps, terminal blocks, and conduit straps o Baffle-former assembly bracket bolts o Thermal shield flexures o Clevis insert bolts o Internals hold-down spring

• Safety and Economic Risk Category C: o Baffle-former assembly comer bolts o Radial support keys (stellite wear surface) o Clevis inserts (stellite wear surface) o Thermal sleeves

Two Category C items from MRP-191, Revision I were reduced to lower risk categories:

• Control rod guide tube (CRGT) assembly C-tubes • Flux thimbles (tubes)



July 9, 2018

The notes with Table 4-10 provide details on why individual components were increased or decreased in risk category. Just like the CE components that were elevated, the revised screening criteria, recent operating experience, and increases in time and neutron dose.influenced the increases in risk category. The decreases in risk category were related to refinement of the relative risk for a component, such as with the comparison between the C-tubes and the sheaths or guide cards and revisions to risk based on actual operational impact, such as for the flux thimble tubes.

This list of high risk category items for SLR, combined with the comparisons made to the previous list of high risk items for the first PEO will be used as a foundation for the expected impacts to MRP-227 when applied to SLR. The components that have elevated safety risks will be treated in Section 5, while those with elevated economic risk will be treated in Section 6. Consistent with the gap analysis requirements provided by the NRC in NUREG-2191 [7] and NUREG-2192 [8], MRP-227-A is assumed to be the starting basis for an SLR application. Section 5 provides changes that increase, clarify, or tighten those requirements along with some new requirerp~nts. Even though it may be possible to reduce the MRP-227-A requirements in some cases, this interim guidance does not attempt to do so. Thus, an SLR application should treat this interim guidance as supplementary to the existing MRP-227-A guidance. The NRC safety evaluation ofMRP-227, which has not yet been completed at the time of issuance of this interim guidance, may allow for relaxation of certain inspection requirements versus what is currently in MRP-227-A. Some of these instances have been identified in notes of the tables provided in Section 5.

Section 6 is provided as good practice guidance for long-term asset management of the reactor internals components.

Note that the MRP-191 SLR expert panel results are currently still in a draft form and have not been finalized by publication of Revision 2 of MRP::.191. The meeting minutes have been completed and verified [6], so the basic expert panel outputs are not expected to change. However, as MRP-.191, Revision 2 is written, verified, and reviewed by the members of the MRP, it is possible that some aspects will be adjusted. Any application of this interim guidance should review the subsequent published dpcuments for impacts on the conclusions provided here.

(


Table 4-7: Combustion Engineering SLR Expert Panel Review Results Table for MRP-191

Assembly/ Screened-in

Likelihood Safety Economic Component Material Degradation

Subassembly Mechanisms

of Failure Consequence Consequence

Upper Weld, IASCC,

Internals Fuel alignment plate 304 ss Wear, Fatigue, M M H

Assembly IE

Lower System 80 Core support Weld, IASCC,

Support · deep beams2 304 ss

Fatigue, IE, VS M M H

Structure

Upper cylinder girth 304 ss Weld, IASCC,

M M H welds (UFW and UGW) Fatigue, IE

Upper cylinder axial 304 ss Weld, IASCC,

M M H welds(UAW) Fatigue, IE

Lower cylinder girth Weld, IASCC,

welds (MGW and 304 ss M M H CSB

LGW/LFW) Fatigue, IE

Assembly Lower cylinder axial Weld, IASCC,

welds (MAW and LAW) 304 ss

Fatigue, IE M M H

Lower core barrel flange 304 ss Weld, Fatigue M L H5

flexure weld3

Core stabilizing lug shim X-750 SCC, Wear H L H6

bolts4

IASCC, Wear,

Core shroud tie rods 348 ss Fatigue (I), IE, ·H1 L1 H1

Core Shroud VS, ISR/IC

Assembly IASCC, Wear,

Core shroud tie rod nuts 316 ss Fatigue (I), IE, H1 L H VS, ISR/IC

Safety Economic

FMECA FMECA Grouo Groun

)

2 3

2 3

2 3

2 3

2 3

2 3

2 3

3 3

3 3

~ 3 . .J


Page 18 of70

L TR-AMLR-18-4, Rev. 1 July 9, 2018

Safety Economic MRP-227-A

Category Category Category

B C P,N1

C C p

B C P,E

B C E

B C p

B C E

B C p

B C N/A4

B C N

B C N

1. 2. 3. 4. 5. 6. 7.

Primary for System 80 style plants (with full-height welded core shroud assembly), No Additional Measures for other plant designs

Name modified to specify that the core support deep beams are only applicable to the System 80 design

Originally included in the lower cylinder of the core support barrel-separated out and named specifically to add clarity

New component added to the MRP-191 list during SLR expert panel review

Consequence could be reduced by development of disposition through analysis

Consequence could be reduced by implementing proactive aging management options

Rankings based on the likelihood and consequence of one tie rod failing

P = MRP-227-A Primary inspection component

E = MRP-227-A Expansion inspection component

N = MRP-227-A No Additional Measures component

NIA= not applicable

T bl 4 8 W f h a e - : es m21 ouse SLRE xpert p ane

Assembly Subassembly Component Material

Flexures2 X-750

- Guide 304 ss \ . ~ plates/cards 316L SS

CRGT1

Guide assemblies and

plates/cards CF8

flow downcomers Guide tube

X-750 Upper support pins

Internals Sheaths 304 ss Assembly

UCP and fuel Upper core 304 ss alienment pins plate

Upper Brackets, clamps,

instrumentation terminal

304 ss conduit and

blocks, and 302 ss4

supports conduit straps

Baffle-edge 316 ss bolts 347 ss

Bracket Bolts5 347 ss

Lower Baffle and Baffle-former 316 ss Internals former

bolts 347 ss Assembly assembly

Corner Bolts6 347 ss

Barrel-former 316 ss bolts 347 ss


fR ev1ew R esu ts T bl ti MRP 191 a e or -Screened-in

Likelihood Safety Economic Degradation Mechanisms

of Failure Consequence Consequence

SCC, Fatigue H L M Weld, Wear,

H H M Fatigue

Weld, Wear, Fatigue, TE

H H M

SCC,IASCC, Wear, Fatigue, H L M

IE, ISR/IC Wear, Fatigue H H M

IASCC, M M H Fatigue, IE

SCC, Fatigue H L H

IASCC, Wear (I), Fatigue, IE, H L M

VS,ISR/IC

IASCC, Wear \

(I), Fatigue, IE, H L M VS,ISR/IC

IASCC, Wear (I), Fatigue, IE, H M M

VS,ISR/IC IASCC, Wear

(I), Fatigue, IE, H M M VS,ISR/IC

IASCC, Wear (I), Fatigue, IE, H L M

VS,ISR/IC

Safety Economic FMECA FMECA Group Group

3 3

3 3

3 3

3 3

3 3

2 3

3 3

3 3

3 3

3 3

3 3

3 3



July 9, 2018

Safety Economic MRP-227-A Category Category Category

C C N2

C C p

C C p

C C X

C C N3

B C E

B C N

B C p

B C p

C C p

C C p

B C E

/


Table 4-8: Westin~house SLR Expert Panel Review Results Table for MRP-191 (cont.) --

Screened-in Likelihood Safety Economic

Assembly Subassembly Component Material Degradation of Failure Consequence Consequence

Mechanisms

Lower core -,

barrel axial Weld, IASCC,

welds (MAW 304 ss Fatigue, IE, M M H

and LAW) vs

Lower core Weld, IASCC,

barrel girth 304 ss Fatigue, IE, M M H

welds (LGW and vs Core Barrel LFW)

Lower Upper core

barrel axial 304 ss Weld, Fatigue M M H Internals

welds(UAW) Assembly

Upper core (cont.)

barrel girth 304 ss Weld, Fatigue M M H

welds (VFW and UGW)

Neutron panels/ Thermal shield 304 ss

Weld, Fatigue M L H thermal flexures 316 ss shield

Radial Radial support Stellite7

• Wear H M M support keys keys

Clevis insert X-750 SCC, Wear H L H

bolts

Interfacing Clevis inserts Stellite Wear, Fatigue H M H

Interfacing Internals hold-

Components components down spring

304 ss Wear, Fatigue9 H L H

Thermal 304 ss _ Wear H 'H H

sleeves10

I. CRGT = control rod guide tube

Safety Economic

FMECA FMECA Group Group

2 3

2 3

2 3

2 3

2 3

3_ 3

3 3

3 3

3 3

3 3



_ July9,2018

MRP-227-Safety Economic

Category Category A

Category

B C E

B C p

B C E

B C p

B C p

C C N

B C X

C C NS

B C p

C C N/AlO

2. The Alloy X-750 CRGT flexures are only applicable to one plant, since the rest have already implemented replacements _

3. The CRGT assembly sheaths, were placed in the No Additional Measures group, because degradation of the sheaths is expected to be bounded by that in the CRGT assembly

guide cards (wear) and lower flanges (cracking) -4. Ranking is bounding for the two material types


5. Previously included with the baffie-edge bolts-separated out tq be improve clarity of the component list and table results 6. Previously included with the baffie-former bolts-separated out to be improve clarity of the component list and table results 7. Material added to the MRP-191 Iist during SLR expert panel review 8. Component added as an Existing item in MRP-227, Revision· I [2] 9. Internals hold-down spring included due to the potential for thermal stress relaxation 10. New component added to the MRP-191 list during SLR expert panel review

P = MRP-227-A Primary inspection component E = MRP-227-A Expansion inspection component _X = MRP-227-A Existing inspection component N = MRP-227-A No Additional Measures component NIA= not applicable


Page21 of70 LTR-AMLR-18-4, Rev. 1

July 9, 2018



July 9, 2018

Table 4-9: Comparison between CE Catee:orv C Components from MRP-191, Revision 1 f5] and the SLR Expert Panel Review

Assembly/Subassembly Component MRP-191, Rev. 1 SLR Expert Panel SLR Expert Panel MRP-227-A

Cate1!0ry Safetv Categorv Economic Category Category

Upper Internals Assembly Fuel Alignment Plate B B ci P,N2

Lower Support Structure System 80 Core support deep beams3 C C C p

Fuel alignment pins (A286SS) C A4 B4 X

Upper cylinder (includes UFW, UGW, and UAW) B BS cs pfE6

Lower cylinder (includes MGW, LGW/LFW, C BS cs P/E6

Core Support Barrel Assembly MAW, LAW, and CSBFW)7

Lower core barrel flange flexure weld7 C B C p

Core stabilizing lug shim bolts8 NIA B C N/A8

Shroud plates C B9 B9 p

Core Shroud Assembly Core shroud tie rods B B cio N

Core shroud tie rod nuts B B cio N

In-core Instrumentation ICI thimble tubes-lower C Bil Bil xll 1. Fuel alignment plate economic risk category increased from MRP-191, Rev. 1 because a non-functional plate would likely require replacement of the full upper internals.

2. Primary for System 80 style plants (with full-height welded core shroud assembly), No Additional Measures for other plant designs

3. Component renamed from "core support deep beams" to "System 80 core support deep beams," since this is only applicable to System 80 plants.

4. Risk category reduced from C for A286 fuel alignment pins because of redundancy and lack of operating experience.

5. The Safety Category for the upper cylinder welds was maintained at B while the Economic Category was increased to C. The Safety Category for the lower cylinder welds

was decreased to B while the Economic Category was maintained at C. The safety category is based on the barrel being able to tolerate some cracking and continue operation.

Based on operating experience showing circumferentially-oriented flaws at an axial weld, the axial welds were assigned the same ~afety and economic levels as the girth welds.

6. The upper core support barrel flange weld, lower cylinder girth welds, and lower flange weld are Primary components in MRP-227-A. The lower core barrel flange, upper

cylinder (including welds), upper core barrel flange, and cote barrel assembly axial welds are Expansion components in MRP-227-A.

7. For SLR expert panel review, the lower cylinder was divided into three separate components: lower cylinder girth welds (MGW and LGW/LFW), lower cylinder axial welds

(MAW and LAW), and lower core support barrel flange flexure weld (CSBFW). The CSBFW stayed in Category C.

8.. The core stabilizing lug shim bolts are new components added during the SLR expert panel review. 1

9. The shroud plates were originally in the highest risk category due to being the lead component for VS. Based on recent research developments on VS [35], the effect of VS is

expected to be significantly smaller, resulting in a reduced risk category assignment. The lack of observed VS in oper~ting plants to date was also considered.

10. The core shroud tie rods were elevated due to the addition ofIASCC as an applicable mechanism during the SLR expert panel. The rods are in tension and the baffie-former

bolt experience has some relevance. Additionally, the economic impact of failed tie rods could be severe. This was further impacted by the Spring 2018 operating experience

ofa tie rod wearing through its threac,ls and falling out of position. Based on these considerations and this operating experience, both the tie rod and the tie rod nut were

elevated to economic risk category C. 11. The risk category of the ICI thimble tubes - lower was reduced based on the fact that all of the affected plants have implemented design changes. These-design changes appear

to have addressed the original issue of excessive radiation growth of the tubes.

P = MRP-227-A Primary inspection component E = MRP-227-A Expansion inspection component X = MRP-227-A Existing inspection component

N = MRP-227-A No Additional Measures component NIA= not applicable

C

Enclosure to MRP 2018-022 Westinghouse Non-Proprietary Class 3


July 9, 2018

Table 4-10: Comparison between Westinghouse Category C Components from MRP-191, Revision 1 [5] and the SLR Expert Panel Review

Assembly Subassembly Component MRP-191, SLR Expert Panel SLR Expert Panel MRP-227-A

Rev. 1 Cate2ory Safety Cate2ory Economic Cate2ory Cate2ory Flexures1

C C C NI

Guide plates/cards C C C p CRGT assemblies and flow

Guide tube support pins (Alloy X-750) C C C X downcomers Upper C tubes C B2 B2 N Internals

Assembly Sheaths C C C N UCP and fuel alignment pins Upper core plate B B CJ -

E Upper instrumentation conduit Brackets, clamps, terminal blocks, and conduit

A B4 c4 N and supports straps.

Baffle-edge bolts C B C p

Bracket bolts5 NIA5 B C pS

Baffle and former assembly Baffle-former bolts C C C p

Comer bolts6 N/A6 C C p6

Lower Barrel-former bolts C B C E

Internals Lower core barrel (includes LGW, LFW, MAW, and C BS cs P/E9

Assembly Core barrel LAW)7

Upper core barrel (includes UFW, UGW, and C BS cs P/E9

UAW)IO

Flux thimbles (tubes) Flux thimbles (tubes) C Bil Bil X Neutron panels/ thermal shield Thermal shield flexures B B c12 p

Radial support keys Radial support keys13 A - c13 c13 N Clevis insert bolts B B c14 X

Interfacing Interfacing components

Clevis inserts A els CIS N components Internals hold-down spring B B c16 p

Thermal sleeves17 NIA C C N/A11

1. C~GT assembly flexures fabricated from Alloy X-750 are only applicable to one plant. 2. The aging degradation mechanism of concern for the CRGT C-tubes is wear. Based on the examinations conducted to date, they are not the lead items for wear in the CRGT

assemblies. Thus, the expert panel assigned them to a lower risk category than in MRP-191, Revision 1. 3. UCP economic risk category increased from MRP-191, Revision 1 because a non-functional plate would likely require replacem~nt of the full upper internals assembly. 4. The brackets, clamps, terminal blocks, and conduit straps were increased from risk Category A to B and C based on the operating experience with cracking of clamps. 5. Previously included with the baffle-edge bolts-separated out to be improve clarity of the component list and table results 6. Previously included with the baffle-former bolts-separated out to be improve clarity of the component list and table results



July 9, 2018

7. For SLR expert panel review the lower core barrel was divided into two separate components: lower core barrel girth welds (LGW and LFW) and lower core barrel axial welds (MAW and LAW).

8. The Safety Category for the lower and upper core barrel welds was reduced to B and the Economic Category was maintained at C. The safety category is based on the fact that

the core support barrel can tolerate some level of cracking, allowing a crack to be dispositioned for continued operation. Based on the operating experience showing

circumferentially-oriented flaws at an axial weld, the axial welds were assigned the same safety and economic levels as the girth welds. 9. The upper core barrel flange weld, upper and lower core barrel cylinder girth welds, and lower core barrel flange weld are Primary components in MRP-227-A. The core

barrel outlet nozzle welds and upper and lower core barrel cylinder axial welds are Expansion components in MRP-227-A. 10. For SLR expert panel review the upper core barrel was divided into two separate components: upper core barrel girth welds (UFW and UGW) and upper core barrel axial

welds (UAW). 11. The flux thimble tubes were reduced from C to B based on the fact that these have a low consequence of failure and they can be relatively easily replaced. Note that the

likelihood of failure and both of the consequence categories remained the same as those determined in MRP-191, Rev. 1.

12. The economic risk category of the thermal shield flexures was elevated to Category C based on the high difficulty ofreplacing or repairing the flexure and the likelihood of

having to remove the thehnal shield to address loss of flexure function. Removing the thermal shield could impact the capability of some plants to operate to 80 years due to

radiation embrittlement of the vessel. 13. The stellite hard-facing surface on the radial support keys was added by the SLR expert panel review. It was elevated to risk category C by the panel due to the potential for

wear by 80 years ofoperation that could result in a loss of function of this alignment component. 14. The economic consequence of the clevis insert bolts was elevated based on the high likelihood of degradation (these have already failed in service) and the potentially high

economic consequence if bolt failures are left unaddressed. 15. The stellite hard-facing surface on the clevis inserts was elevated to risk category C by the panel due to the potential for wear by 80 years of operation that could result in a loss

of function of this alignment component. 16. Type 304SS hold-down springs are susceptible to thermal stress relaxation. The SLR expert panel elevated these hold-down springs to economic risk category C because of

the multiple other components that can be affected by a relaxed hold-down spring. The economic impact would be much less if this component is addressed proactively.

17. The thermal sleeves were added by the SLR expert panel in response to the known operating experience [29] [30] [31] [32].

P = MRP-227-A Primary inspection component E = MRP-227-A Expansion inspection component X = MRP-227-A Existing inspection component N = MRP-227-A No Additional Measures component NIA= not applicable


5 Expected impacts to MRP-227 for SLR



July 9, 2018

A summary of the Primary, Expansion, and Existing Components from MRP-227-A [1] is provided in the following tables: '

• CE Primary and Expansion Components: Table 5-1 • CE Existing Components: Table 5-2 • Westinghouse Primary and Expansion Components: Table 5-3 • Westinghouse Existing Components: Table 5-4

For reference and information, the same is provided for the MRP-227, Revision 1 [2] components in the subsequent tables:

• CE Primary and Expansion Components: Table 5-5 • CE Existing Components: Table 5-6 , • Westinghouse Primary arid Expansion Components: Table 5-7 • Westinghouse Existing Components: Table 5-8

Developing interim guidance on the likely impacts to MRP-227 for SLR and supporting the gap analysis between MRP-227-A and the guidance needed for SLR requires answers to the following three questions:

• What changes need to be made to the current Primary, Expansion, or Existing inspection requirements? Including the following details:

o Timing o Inspection type o Coverage o Degradation mechanism of concern

• What additional components need to be added to the Primary, Expansion, or Existing categories for SLR? Including the following details:

o Timing o Inspection type o Coverage o Degradation mechanism of concern

• What components can be removed from the MRP-227 requirements based on updated knowledge?

o Only components which are no longer applicable due to plant shutdowns or imminent plant shutdowns have been removed from the requirements

The interim guidance contained in this section was developed with consideration of these three questions in light of the safety consequence and category results for the SLR expert panel review presented in Section 4. High economic risk components will be addressed in Section 6 to support effective asset management of the internals.

Consistent with the NRC requireinents for the SLR gap analysis, the additions and changes provided here are based on an assumption that MRP-227-A is the starting point for an SLR application. The industry



July 9, 2018

has already provided the first step in performing this SLR gap analysis through publication of the updated

I&E Guidelines in MRP-227, Revision 1. MRP-227, Revision 1 documented the latest requirements

based on current industry knowledge at the time of its publication. In the following sections, the

differences between MRP-227-A and MRP-227, Revision 1 will be reviewed first. Then, the additions

and changes based on the SLR expert panel review outputs will be provided.

Impacts on future MRP-227 guidance due to other interim guidance or responses to RAis will also be

. addressed in this section. The evaluations of this section will be based primarily on the elevated safety

risk items listed in Section 4. These elevated safety risk category components are the most likely to rise

from a lower inspection requirement category to a higher inspection category for SLR, particularly if the

results for that component changed significantly.

Inspection timing for items. in the second PEO may be complicated by the inspection timing at a plant for

the first PEO. The following are guiding points to clarify the required timing:

• Any component that is already being inspected during the first PEO using the specified re

inspection interval, which does not receive changes to inspection type or coverage, will continue

to be inspected into the second PEO using that same re-inspection interval, unless otherwise

specified. o Example 1: First PEO Primary, Expansion, or Existing component that will not be

changing category or inspection type or coverage from first PEO to second PEO. If the

component was last inspected in the first PEO at 57 years with a 10-year re-inspection

interval, it will perform the next inspection at approximately 67 years and not at 60 years.

o Example 2: First PEO Expansion, or Existing component that will be changing category

from first PEO to second PEO but the inspection type or coverage will not change. Same

as Example 1, if the component was last inspected in the first PEO at 57 years with a 10-

year re-inspection interval, it will perform the next inspection at approximately 67 years

and not at 60 years. (Example: clevis insert bolts moving from Existing to Primary)

• Any component that is already being inspected during the first PEO using the specified re

inspection interval, which does receive changes to inspection type or coverage, will be required to

implement the new inspection according to the initial timing provided for SLR. These will

typically have a range built in to allow planning flexibility. (e.g., "within 2 refueling outages from

the start of the SLR period") o Example 3: First PEO Primary, Expansion, or Existing component that will be changing

inspection type or coverage from first PEO to second PEO. In this case because the

inspection requirements have changed, the new inspection requirements must be

implemented at the required initial timing for second PEO.

• Any component that is not being inspected during the first PEO and is elevated to a Primary or

· Existing inspection category will be required to implement the inspection according to the initial

timing provided for SLR.

As with MRP-227-A, the inspection intervals were typically set to 10 years in order to allow coordination

with existing inspections under·the ASME Code. This allows a utility to minimize the number ofhigh

cost operations, such as core barrel pulls, which must be performed for aging management.


Table 5-1: CE MRP-227-A Primary and Expansion Components (1]

• =Primary o = Expansion

• Core Shroud Assembly (Bolted) - Core shroud bolts o Core Shroud Assembly (Bolted) - Core support column bolts o Core Shroud Assembly (Bolted) - Barrel-shroud bolts

Page 27 of70 LTR-AMLR-18-4, Rev.-1

July 9, 2018

• Core Shroud Assembly (Welded)- Core shroud plate-former plate weld (core shrouds assembled in two vertical sections)

o Core Shroud Assembly (Welded) - Remaining axial welds • Core Shroud Assembly (Welded)- Shroud plates (core shrouds with full-height shrou~ plates)

o Core Shroud Assembly (Welded)- Remaining axial welds o Core Shroud Assembly (Welded) -Ribs and rings

• Core Shroud Assembly (Bolted) - Assembly - · • Core Shroud Assembly (Welded) -Assembly • Core Support Barrel Assembly ~ Upper ( core support barrel) flange weld

o Core Support Barrel Assembly - Lower core barrel flange o Core Support Barrel Assembly- Upper cylinder (including welds) o Core Support Barrel Assembly - Core barrel assembly axial welds o Core Support Barrel Assembly - Upper core barrel flange o Lower Support Structure - Lower core support beams

• Core Support Barrel Assembly - Lower cylinder girth welds o Core Support Barrel Assembly - Lower cylinder axial welds

• Lower Support Structure - Core support column welds • Core Support Barrel Assembly - Lower flange weld • Lower Support Structure - Core support plate • Upper Internals Assembly- Fuel alignment plate (core shrouds with full-height shroud plates) • Control Element Assembly (CEA) - Instrument guide tubes

o Control Element Assembly - Remaining instrument guide tubes within the CEA shroud assemblies

• Lower Support Structure - Deep beams

Table 5-2: CE MRP-227-A Existing Components (l]

Core Shroud Assembly - Guide lugs Core Shroud Assembly - Guide lug inserts and bolts Lower Support Structure - Fuel alignment pins Core Barrel Assembly - Upper flange



July 9, 2018

Table 5-3: Westinghouse MRP-227-A Primary and Expansion Components [1]

• = Primary o = Expansion

• Control Rod Guide Tube Assembly - Guide plates (cards)

• Control Rod Guide Tube Assembly - Lower flange welds

o Upper Internals Assembly - Upper core plate

o Lower Internals Assembly - Lower support forging or castings

o Lower Support Assembly-Lower support column bodies (cast)

o Bottom Mounted Instrumentation System - Bottom-mounted instrumentation (BMI)

column bodies • Core Barrel Assembly - Upper core barrel flange weld

o Core Barrel Assembly - Core barrel outlet nozzle welds

o Lower Support Assembly - Lower support column bodies (non-cast)

• Core Barrel Assembly - Upper and lower core barrel cylinder girth welds

o Core Barrel Assembly - Upper and lower core barrel cylinder axial welds

• Core Barrel Assembly - Lower core barrel flange weld

• Baffle-Former Assembly - Baffle-edge bolts

• Baffle-Former Assembly- Baffle-former bolts o Core Barrel Assembly - Barrel-former bolts

o Lower Support Assembly - Lower support column bolts

• Baffle-Former Assembly - Assembly (Includes: Baffle plates, baffle edge bolts, and indirect

effects of void swelling in former plates) • Alignment and Interfacing Components - Internals hold down spring

• Thermal Shield Assembly - Thermal shield flexures

Table 5-4: Westinghouse MRP-227-A Existing Components [1]

Core Barrel Assembly - Core barrel flange

Upper Internals Assembly - Upper support ring or skirt

Lower Internals Assembly - Lower core plate, XL lower core plate

Bottom Mounted Instrumentation System - Flux thimble tubes

Alignment and Interfacing Components - Clevis insert bolts

Alignment and Interfacing Components - Upper core plate alignment pins



July 9, 2018

Table 5-5: CE MRP-227, Revision 1 Primary and Expansion Components [2]

• = Primary o = Expan_sion

• C 1. Core Shroud Assembly (Bolted) - Core shroud bolts o C 1.1. Core Shroud Assembly (Bolted) - Core support column bolts o C 1.2. Core Shroud Assembly (Bolted) - Barrel-shroud bolts

• C2. Core Shroud Assembly (Welded)- Core shroud plate-former plate weld o C2.1. Core Shroud Assembly (Welded) - Remaining axial welds

• C3. Core Shroud Assembly (Welded)- Shroud plates o C3 .1. Core Shroud Assembly (W eld_ed) - Remaining axial welds o C3.2. Core Shroud Assembly (Welded)-Ribs and rings

• C4. Core Shroud Assembly (Bolted) - Assembly • C4a. Core Shroud Assembly (Welded) -Assembly • CS. Core Support Barrel Assembly- Upper flange weld (UFW)

o CS. I. Core Support Barrel Assembly - Lower Girth Weld (LGW) o CS.2. Core Support Barrel Assembly- Upper Girth Weld (UGW) o CS.3. Core Support Barrel Assembly- Upper Axial Weld (UAW) o CS.4. Lower Support Structure- Lower core support beams

• C6. Core Support Barrel Assembly- Middle Girth Weld (MGW) o C6. l. Core Support Barrel Assembly - Middle Axial Weld (MAW) o C6.2. Core Support Barrel Assembly- Lower Axial Weld (LAW)

• C7. Core Support Barrel Assembly- CSB Flexure Weld (CSBFW) • CS. Lower Support Structure - Core support columns • C9. Lower Support Structure - Core support plate • C 10. Upper Internals Assembly - Fuel alignment plate • C 11. Control Element Assembly - Instrument guide tubes

o C 11.1. Control Element Assembly - Remaining instrument guide tubes • C12. Lower Support Structure - Deep beams

Table 5-6: CE MRP-227, Revision 1 Existing Components [2]

Cl3. Core Shroud Assembly-Guide lugs C 14. Upper Internals Assembly - Guide lug inserts and bolts Cl Sa. and Cl Sb. Lower Support Structure- Fuel alignment pins Cl 6. Core Barrel Assembly- Upper flange C 17. Alignment and Interfacing Components - Core Stabilizing Lugs and Shims



July 9, 2018

Table 5-7: Westinghouse MRP-227, Revision 1 Primary and Expansion Components [2]

• =Primary o = Expansion

• Wl. Control Rod Guide Tube Assembly - Guide plates (cards) • W2. Control Rod Guide Tube Assembly - Lower flange welds

o W2. l. Control Rod Guide Tube Assembly - Remaining CRGT lower flange welds

o W2.2. Bottom Mounted Instrumentation System - Bottom-mounted instrumentation

(BMI) column bodies

• W3.Core Barrel Assembly- Upper flange Weld (UFW) o W3 .1. Core Barrel Assembly - Upper Girth Weld (UGW)

o W3.2. Core Barrel Assembly- Upper Axial Weld (UAW)

o W3.3. Core Barrel Assembly- Lower Flange Weld (LFW)

o W3.4. Lower Internals Assembly- Lower support forging or castings

• W 4. Core Barrel Assembly - Lower girth weld (LGW) o W 4.1. Upper Internals Assembly - Upper core plate o W4.2. Core Barrel Assembly- Middle Axial Welds (MAW)

o W4.3. Core Barrel Assembly- Lower Axial Welds (LAW) o W4.4. Lower Support Assembly- Lower support column bodies (both cast and non-cast)

• WS. Baffle-Former Assembly- Baffle-edge bolts • W6. Baffle-Former Assembly- Baffle-former bolts

o W6. l. Core Barrel Assembly- Barrel-former bolts o W6.2. Lower Support Assembly - Lower support column bolts

• W7. Baffle-Former Assembly-Assembly (Includes: Baffle plates, baffle edge bolts, corner bolts,

and indirect effects of void swelling in former plates)

• W8. Alignment and Interfacing Components - Internals hold down spring • W9. Thermal Shield Assembly- Thermal shield flexures

Table 5-8: Westinghouse MRP-227, Revision 1 Existing Components [2]

Wl 0. Core Barrel Assembly - Core barrel flange Wl 1. Upper Internals Assembly - Upper support ring or skirt W12a and W12b. Lower Internals Assembly- Lower core plate, XL lower core plate

W13. Bottom Mo~ted Instrumentation System - Flux thimble tubes Wl4. Alignment and Interfacing Components - Clevis bearing Stellite wear surface

Wl 4. Alignment and Interfacing Components - Clevis insert bolts WlS. Alignment and Interfacing Components - Upper core plate alignment pins

I Enclosure to MRP 2018-022 Westinghouse Non-Proprietary Class 3


July 9, 2018

5.1 Impacts on CE Inspection Requirements

5.1.1 Differences between MRP-227-A and MRP-227, Revision 1 Inspection Requirements

This section addresses the differences between the baseline MRP-227-A l&E requirements and the revised MRP-227, Revision I requirements. The following list of changes to the MRP-227 inspection program tables were a mix of more conservative criteria, less conservative criteria, and clarifications:

• Changes for Primary components o Clarifications on the examination coverage for the VT-I inspection of the welded core

shroud assembly for plant designs with core shrouds assembled in two vertical sections. o Clarific~tion on the naming of the upper flange weld (UFW) o Changes to the expansion components for the UFW

• Expan~ion to upper cylinder welds increased to include all low fluence girth welds and axial welds (UGW, LGW/LFW, and UAW) ·

• Expansion to upper core barrel flange removed o Lower core barrel cylinder girth welds narrowed from all welds to the middle girth weld

(MGW) as the lead weld • MA w_ and LAW set as the Expansion components from the MGW

o Reduction of inspection coverage for Primary core barrel welds to 2?% o Reduction of inspection coverage for core support columns to a VT -3 of 25% of the

column assemblies for plants with bolted core shrouds and '100% of the column welds from the top side of the core. support plate for plants with welded core shrouds assembled in two vertical sections.

o Naming of the core support barrel flexure weld (previously "lower flange weld") clarified and SCC added to the applicable degradation mechanisms

o IE added as a degradation mechanism for the fuel alignment plate o Examination coverage for the lower support structure deep beams reduced to 25% of the

total number of beam-to-beam welds • Changes for Expansion components

o Naming of lower girth weld (previously "lower core barrel flange") clarified o Examination coverage for the LGW, UGW, UAW, MAW, LAW, and remaining core

shroud assembly axial welds changed from I 00% of accessible welds with a minimum of 75% coverage to a minimum of75% of the weld surface

o Naming of core support barrel upper cylinder clarified to be the upper girth weld (UGW) and upper axial weld (UAW)

o Component upper core barrel flange removed o Naming of component core barrel axial welds clarified by placing the UAW with the

welds expanding from the UFW and by placing the MAW and LAW with the welds expanding from the MGW

o Coverage of lower core support beams reduced to 25% o Expansion component core shroud assembly (welded) - remaining axial welds added to

clarify the expansion from the core shroud plate-former plate weld for plant designs with core shrouds assembled in two vertical sections



July 9, 2018

o Expansion component entry containing remaining axial welds and ribs and rings clarified to just apply to plants with welded core shrouds assembled with full-height core shroud

plates o Inspection coverage for the ribs and rings set to 25%

• Changes for Existing components o Added the core stabilizing lugs and shims

For the gap analysis with MRP-227-A, only the more conservative or improved criteria will be relevant to

an SLR application. The more conservative criteria from the above list are: ,

• Addition of SCC to the applicable mechanisms for the core support barrel flexure weld

(previously "lower flange weld") • Addition of IE as a degradation mechanism for the fuel alignment plate

• Addition of the Expansion component core shroud assembly (welded)- remaining axial welds

• Addition of the core stabilizing lugs and shims

Adding these items from MRP-227, Revision 1 to the SLR aging management program is the first step in

meeting the requirements of the SLR gap analysis.

For the remaining changes, including the clarifications should be considered for the improved

understanding that they provide. The addition of the less conservative criteria will require inclusion of an

adequate technical basis. Much of this technical basis has already been developed through the MRP-227,

Revision 1 basis documents and responses to NRC RAis, but use of the less conservative criteria carries

an inherent risk since MRP-227, Revision 1 has not yet received an NRC safety evaluation.

5.1.2 Changes to the Current Primary, Expansion, or Existing inspection Requirements

The SLR expert panel review did not result in any changed requirements for the current Primary,

Expansion, or Existing inspection items in MRP-227-A [1] and MRP-227, Revision 1 [2], based on the

safety ris~ categorization. Five CE components or component groups may require changes from the guidance in MRP-227, Revision 1 based on consideration of the NRC RAis on MRP-227, Revision 1 and

the responses to those RAis:

• System 80 core support deep beams • Lower core barrel flange flexure weld • Core support barrel welds • Core support columns • Welded core shroud assembly ribs and rings

The details of the potential revisions to the inspection requirements of these five components are provided in greater detail in the RAI response letters [14] [15] [16]. These letters include the NRC RAls, the detailed responses, and proposed revised entries for the MRP-227, Revision 1 tables . .A summary of these

changes is provided here:


• System 80 core support deep beams



July 9, 2018

o The response to RAI 10 clarified the inspection coverage of the welds by specifying that the25% coverage of the deep beam structure must be evenly distributed across the component assembly.

• Lower core barrel flange flexure weld o The response for RAI 16 clarifies that if the lower core barrel flange flexure weld (also

known as core support barrel flexure weld or CSBFW) screens out for fatigue, SCC must still be considered. This can be accomplished by performing an evaluation using plantspecific or bounding information, or by performing the prescribed inspection.

• 1 Core support barrel welds o The response to RAI 5 provided a basis for the reduced inspection coverage of the core

support barrel welds, and RAI 26 provided a basis for the assignment of welds to the Primary or Expansion component lists. These changes were implemented in MRP-227, Revision 1 but have not yet been accepted by the NRC staff. Note that the recerit operating experience finding crack-like indications in a CE-design core support barrel may have some impact on the final requirements for these welds. This interim guidance does not create revised requirements for CE core support barrel welds, since the industry is currently evaluating the need for further guidance on core barrel aging management.

• Core support columns o The response to RAI 9 states that the core support columns can be moved from a Primary

inspection component to an Expansion inspection component. • Welded core should assembly ribs and rings

o RAI 12 questioned whether these components were accessible, and the response to the question concluded that they were not accessible given currently available inspection techniques. The proposed change to address this was to revise the "Examination Method/Frequency" to be "No exarp.ination requirements. Justify by evaluation or by replacement."

5.1.3 Additional Components to be added to the Primary, Expansion, or Existing Categories It is expected that three new components will be added to or elevated within the MRP-227-A or MRP-227, Revision 1 CE Primary, Expansion, and Existing component inspection requirements, based on the MRP-191 SLR expert panel results:

• Primary · o Core stabilizing lug shim bolts o Core shroud tie rods

• Expansion o Fuel alignment plate for all CE plants other than System 80 designs

• Existing o None


Core stabilizing lug shim bolts



July 9, 2018

The experi~nce in Westinghouse plants with clevis insert bolt cracking [33] is also relevant for the CEdesigned plants. As noted in Section 3, this is due to the susceptibility of the Alloy X-750 bolts to primary water SCC. The CE core stabilizing lug shim bolts perform essentially the same function as the Westinghouse clevis insert bolts and were fabricated from the same susceptible material. Therefore, the SLR expert panel added the core stabilizing lug shim bolts to the scope of MRP-191 and assigned them to safety risk C~tegory Band economic risk Category C [6].

The core stabilizing lugs and shims were not included in MRP-227-A and were added in MRP-227, Revision 1 as Existing inspection items (see Section 5.1.1). The general visual (VT-3) ASME Code Secti,:m XI inspection that was specified for the lugs and shims may detect degradation in the bolts, but it also may not. The core stabilizing lug shim bolts could be added as a specific component to be inspected by this ASME Code Section XI VT-3; however, the fact that a similar component for Westinghouse plants has already shown evidence of degradation makes it prudent to increase the requirement for this component:

No other component in the CE reactor vessel internals could serve as a leading indicator for the primary water SCC that the core stabilizing lug bolts are expected to experience. This, combined with the relevant operating experience, supports adding the bolts to MRP-227, Revision 2 as a Primary Component. Due to the fact that degradation has already been observed in a similar Westinghouse component before plants reach the first PEO, it is recommended that the initial inspection be set for the start of the first PEO at 40 years. Consistent with other components, the re-inspection interval w~uld be a maximum of 10 years, but re-inspection may be necessary whenever the core barrel is removed from the vessel, based on the economic risk associated with bolt degradation.

TypicaUy, the safety risk Category B assignment of the core stabilizing lug bolts would not automatically result in the bolts being assigned a Primary inspection. However, the multiple instances of active degradation in an equivalent component and the potential for varying Existing component inspection requirements from plant-to-plant merit a more conservative aging management approach.

A general visual (VT~3) inspection is the current MRP-227-A and MRP-227, Revision 1 requirement for the analogous component, the clevis insert bolts. However, this visual examination will only detect degradation once it has proceeded to the point of complete fracture of the core stabilizing lug bolts and wear on the locking devices and core stabilizing lug shim. Evaluations of potential core stabilizing lug shim bolt degradation were documented in PWROG-15034-P [38]. This study determined that degraded core stabilizing lug shim bolts would not result in a loss of function of the lower radial support system and thus should not pose a safety concern. Several physical constraints were shown to prevent this loss of function during operation. This technical basis supports using a general visual (VT-3) examination to manage the safety function of the core stabilizing lug shim bolts. Proactive replacement of these bolts would also be an acceptabl~ approach to managing the potential aging degradation.

The degradation mechanism is equally likely to occur at any location, so the required coverage should be 100% of the accessible core stabilizing lug bolts. All of the bolts are expected to be accessible.



July 9, 2018

The revised inspection requirements adding the core stabilizing lug bolts to MRP-227, Revision 2 Table 4-2 are shown in Table 5-9. This table provides Primary component additions that should be implemented during the first PEO at 40 years of operation. The guidance in this table constitu~es an ~I 03-08 "Needed" requirement.

Core shroud tie rods

The core shroud tie rods are only applicable to some of the CE plants designed with welded core shrouds assembled in two vertical sections and are not applicable to CE plants designed with full-height core shrouds. The degradation mechanisms of concern for the tie rods are ~SCC, wear, fatigue, IE, VS, and ISR/IC [22]. The wear and fatigue are concerns because of the irradiation-induced stress relaxation that will occur for all tie rod locations. The rods are bolt-like and could be subject to IASCC issues like other irradiated bolts in PWR primary water.

The tie rods are mostly inaccessible, with only the top end being easily visible where it comes through the top former of the core shroud. The SLR expert panel was concerned that the likelihood of IASCC cracking of the tie rods could be high within an 80-year lifetime based on similar relevant experience for bolts. However, the recent operating experience at a CE-designed plant has shown that the tie rods may degrade by other mechanisms, since that tie rod was displaced downwards and showed evidence of wear. The lack of accessibility will make it very difficult or perhaps impossible to replace or repair the tie rods.

The actual operating experience for the core shroud tie rods demonstrates that they are lead items for the degradation mechanism or mechanisms that caused the observed experience. At this time, a causal analysis has not been published for the tie rods, so firm conclusions cannot be provided here. It should be noted that the rods are not the most highly irradiated components in the CE internals, Which should impact their susceptibility to irradiation-related degradation effects. They are also not expected to be lead items for fatigue. However, the operating experience for the tie rods supports elevation of the rods to a primary component. Note that alternative methods of confirming the functionality of the core shroud tie rods, such as using neutron noise to monitor the position of the core shroud assembly segments, could also be used to support this primary inspection requirement and manage the degradation.

Degradation in the tie rods has already been observed as the rod dropping from its design position and falling until it is caught by the geometry of the rod. Degradation was also postulated to potentially appear as shifting of the core shroud sections relative to one another or relative to the core support plate; though, this postulated scenario was not evident in the one instance of tie rod degradation observed to date. The current primary inspection requirement of the welded core shroud assembly (CE Primary inspection in Table 5-1 and Table 5-5) would detect this type of shifting or separation. Most of the core shroud tie rod length is concealed within the core shroud assembly and would be inaccessible. The top end of the tie rod is visible on the top of the top former plate of the welded core shroud assembly and the bottom end of the rod is visible below the core support plate. Based on the degradation observed to date, a visual VT-3 inspection of the top end of the core shroud tie rod is effective at detecting the degradation actually experienced. This primary inspection would only be applicable to those plants with core shroud tie rods and would apply to all of the rods in each of those plants (i.e., 100% coverage).



July 9, 2018

Since degradation has already been observed ata plant close to the beginning of the first PEO, the initial

timing of primary inspection requirement for the core shroud tie rods should start at the beginning of the

first PEO. The rate of the observed degradation is unknown and experience has shown that the rods can change from being visually intact to degraded over the course of a single cycle. Therefore, the top end of

the core shroud tfo rods should be inspected by a visual (VT-3) general condition inspection during every

outage. This inspection should include both top down and side views of the top end of each tie rod in

order to confirm that the rods have not dropped or shifted.

The revised inspection requirements adding the core shroud tie rods to MRP-227, Revision i Table 4-2

are shown in Table 5-9. This also requires some modification of the entry for the welded core shroud in

MRP-227, Revision 2 Table 4-2, which is shown in Table 5-11, along with the previous MRP-227-A

entry for comparison. The entry in Table 5-11 includes augmentation of the welded core shroud

inspection to include a visual inspection of the seam between the lower core shroud segment and the core

support plate. The guidance in these tables constitutes an NEI 03-08 "Needed" requirement.

Fuel Alignment Plate

The fuel alignment plate was originally only included as an MRP-227-A and MRP-227, Revision 1

Primary inspection component for the System 80 design (plant design with full-height core shroud plates).

This was driven by the potential for fatigue in the component [2]. The other fuel alignment plate designs

were assigned to the No Additional Measures category. Fatigue is still a leading concern for the fuel

alignment plate and is applicable for all of the CE plant designs. However, operation to 80 years

increases the fluence exposure for the plate, resulting in IASCC being applicable for the fuel alignment plate in addition to the degradation mechanisms from MRP-191, Revision 1 [22]. Thus, the fuel

alignment plate screened in for SCC (weld), wear, fatigue, IASCC, and IE.

The addition of IASCC to the list of aging degradation mechanisms affecting the fuel alignment plate,

combined with the expected stress levels on the plate, the continued concern for fatigue, and the function

of the plate to directly hold down the fuel, resulted in the higher economic risk category assigned to the

plate. The safety risk category was maintained at B by the SLR expert panel due to the damage tolerance

of the component (a large network of cracks across multiple ligaments would be required).

Based on the risk category assigned to this component and the increase in the degradation mechanisms

that can affect it, for MRP-227, Revision 2, the fuel alignment plate is expected to be elevated to an

Expansion inspection component for CE plants with welded core shrouds assembled in two vertical

sections. The timing for the initial applicability of this Expansion requirement would be at the start of the /

second PEO. This is due to the concern for increased radiation' effects occurring during the second PEO.

The re-inspection interval would be 10 years, consistent with the other Expansion inspection items with a

potential for degradation but no negative operating experience.

The degradation effect of concern for the fuel alignment plate in any CE plant is cracking, whether due to fatigue or IASCC. The fuel alignment plate is similar in geometry, function, and degradation mechanisms to the Westinghouse UCP, which requires an enhanced visual (EVT-1) examination of 100% of the 1

accessible surfaces [1]. This same inspection requirement is appropriate for the Expansion inspection of the fuel alignment plate. This 100% EVT-1 inspection for the UCP in MRP-227-A has been reduced to a



July9; 2018

general visual (VT-3) of 25% of the cor~-side surfaces of the plate for MRP-227, Revision 1 [2]. This was based on considerations of the geometry and functionality of the UCP [14]. The same considerations would also apply to the fuel alignment plate; however, MRP-227, Revision 1 is still under NRC review for a safety evaluation and these revised inspection requirements have not yet been approved. For thjs interim guidance, the inspection requirements of the current approved version and starting point of the gap analysis will be specified. Upon completion ofthe approved version ofMRP-227, Revision 1, this fuel alignment plate inspection method and coverage can be replaced with the resulting inspection method and coverage for the UCP.

The new inspection requirement creating a fuel alignment plate Expansion inspection item applicable to all CE plants with welded core shrouds assembled in two vertical sections for MRP-227, Revision 2 Table 4-2 is shown in Table 5-10. This also requires some modification of the entry for the Core Support Barrel Assembly- Lower cylinder girth welds in MRP~227, Revision 2 T~ble 4~2, which is shown in Table 5-11, along with the previous MRP-227-A entry for comparison. The revised examination acceptance and expansion criteria for the lower cylinder girth welds (MRP-227, Revision 2 Table 5-2) are shown in Table 5-12, along with the previous MRP-227-A entry for comparison. The guidance in these tables

, constitutes an NEI 03-08 "Needed" requirement.

5.1.4 Components to be removed from the MRP-227 Requirements

Several components will be removed from the MRP-227, Revision 2 requirements. These are all due to plants that have shut down as of the publication of this interim guidance or are planning to shut down prior to entering the second PEO and have unique design aspects that are not present in any remaining operating plants. Four components will be removed from the CE Primary, Expansion, and Existing Component lists for this reason:

• Primary o Core shroud bolts o Bolted core shroud assembly

• Expansion o Core support column bolts o Barrel-shroud bolts

•· Existing o None

It is expected that no other components will be removed from the CE Primary, Expansion, or Existing component lists for MRP-227, Revision.2.

5.1.5 Elevated Risk Category Items.that are not Expected to Change

There are two components that received Category C risk rankings in MRP-191, Revision 1 [ 5] and lower risk rankings from the SLR expert panel which are not expected to change for MRP-227, Revision 2.

The core shroud plates received a reduced risk ranking due to the reduction in expected VS [35]. However, even with this reduction the plate~ are the lead item for the VS degradation mechanism. Thus, the core shroud plates will remain a Primary inspection component.



July 9, :2018

The fuel alignment pins received a reduced risk ranking due to redundancy and lack of negative operating

experience. They were originally included as an existing component, so removing the pins from

MRP-227, Revision 2 would not reduce inspection requirements. Additionally, the fuel alignment pins -

were not reduced to a No Additional Measures level. Thus, they will be left in MRP-227, Revision 2 as

an Existing inspection component.

There were also a number of components which were elevated from FMECA category A to safety risk

category B by the SLR expert panel, which do not require additional aging management requirements for

SLR:

• CEA shroud bolts o These were elevated to Category B based on the higher radiation dose to the fuel

alignment plate during SLR operation, the known experience with IASCC degradation of

bolts, and the more conservative FMECA group assignment table. The CEA shroud bolt

heads ~re visible on the underside of the fuel alignment plate and have been inspected

regularly during in-service inspections of the plate. This, combined with the redundancy

in the component and the relatively low rankings, is a reasonable basis for leaving these

bolts in the No Additional Measures category.

• System 80 ICI nozzles J o The main degradation mechanism that the expert panel thought would be an issue for

these nozzles was wear due to the interaction with the instrumentation. The likelihood of

failure and consequence result was the same for MRP-191, Revision 1 [5] and the SLR

expert panel review. The increase in the FMECA group was due to the more

conservative FMECA table. The lack of change from MRP-191, Revision 1 and the

relatively low likelihood and consequence rankings support leaving this component in the

No Additional Measures category.

/



July 9, 2018

Table 5-9: Expected New Entries for Table 4-2 for MRP-227, Revision 2 - CE Primary Components for Implementation for First PEO (NEI 03-08 Needed)

Primary Item Applicability Effect (Mechanism) Expansion Link Examination Method/ Ex~mination Coverage Frequency

Core Shroud Assembly All plants with core Cracking (IASCC, Welded Core Shroud Visual (VT-3) Examination of the top (Welded) shroud tie rods Fatigue), Loss of Assembly examination no later than end of I 00% of the core Core shroud tie rods Material (Wear), 2 refueling outages from shroud tie rods. Top-

Distortion (Void the beginning of the first down and side views of Swelling) Aging license renewal period. each tie rod must be management (IE and Subsequent examinations obtained. JSR/IC) during every refueling

outal!;e. Alignment and All plants Cracking (SCC), Loss of None Visual (VT-3) All core stabilizing lug , Interfacing material (Wear) examination no later than shim bolts Components 2 refueling outages from Core stabilizing lug shim the beginning of the first bolts license renewal period.

Subsequent examinations. on a ten-year interval.

Table 5-10: Expected New Entries for Table 4-5 for MRP-227, Revision 2- CE Expansion Components (NEI 03-08 Needed) Expansion Item Applicability Effect (Mechanism) Primary Link Examination MetI,od/ Examination Coverage

Frequency Upper Internals All plants with welded Cracking (IASCC, Core Support Barrel Enhanced visual I 00% of accessible Assembly core shrouds assembled Fatigue), Loss of Assembly: Lower (EVT-1) examination. surfaces.* Fuel alignment plate in two vertical sections material (Wear), Aging cylinder girth welds Subsequent examination (Expansion only management (IE) on a ten-year interval.* applicable after entering SLR period) * The inspection requirement for the fuel alignment plate is analogous to the current requirement for the Westinghouse-design UCP in MRP-227-A [1]. The inspection technique requirement for the UCP was reduced in MRP-227, Revision 1 [2] and justified in responses to RAis on MRP-227, Revision 1 [14]. Once the NRC safety evaluation is complete, the resulting reduced inspection requirements of the Westinghouse-design UCP can be substituted here for the fuel alignment plate.


Westinghouse Non-Proprietary Class 3 Page 40 of70

LTR-AMLR-18-4, Rev. 1 July 9, 2018

Table 5-11: Expected Revised Entries for Table 4-2 for MRP-227, Revision 2- CE Primary Components for Implementation in SLR

(NEI 03-08 Needed)

Current Table Entries (MRP-227-A Table 4-2) I

Primary Item Applicability Effect (Mechanism) Expansion Link Examination Method/ Examination Coverage Frequency

Core Shroud Plant designs Distortion None Visual (VT-1) examination no If a gap exists, make three to

Assembly with core (Void Swelling), as evidenced by

later than 2 refueling outages five measurements of gap

(Welded) shrouds separation between the upper and

from the beginning of the license opening from the core side at


lower core shroud segments renewal period. Subsequent the core shroud re-entrant

two vertical examinations on a ten-year corners. Then, evaluate the

sections Aging Ma!lagement (IE) interval. swelling on a plant-specific basis to determine frequency and method for additional requirements.

Core Support All plants Cracking (SCC, IASCC) Lower cylinder axial welds Enhanced visual (EVT-1) 100% of the accessible

Barrel Aging Management (IE) examination no later than 2 surfaces of the lower cylinder

Assembly refueling outages from the welds.

Lower cylinder - beginning of the license renewal

girth welds period. Subsequent examinations on a ten-year interval.



July 9, 2018

Table 5-11: Expected Revised Entries for Table 4-2 for MRP-227, Revision 2 - CE Primary Components for Implementation in SLR (NEI 03-08 Needed) {cont.)

Revised Table Entries Pri~ary Item Applicability Effect (Mechanism) Expansion Link Examination Method/ Examination Coverage

Frequency

Core Shroud Plant designs Distortion None Visual (VT-1) examination no 100% of the horizontal seam Assembly with core

(Void Swelling), as evidenced by later than 2 refueling outages between the upper and lower

(Welded) shrouds measurable separation between

from the beginning of the first core shroud segments.


the upper and lower core shroud license renewal period.

100% of the seam between two vertical segments or by shifting of the

Subsequent examinations on a the lower core shroud sections ten-year interv1:1L '-- ' segments relative to one another

' segment and the core support

or the core support plate plate

Aging Management (IE)

Core Support All plants Cracking (SCC, IASCC) Aging Lower cylinder axial welds Enhanced visual (EVT-1) 100% of the accessible Barrel Management (IE)

Fuel alignment plate (Plant examination no later than 2 surfaces of the lower cylinder

Assembly refueling outages from the welds. designs with core shrouds

beginning of the first license Lower cylinder assembled in two vertical renewal period. Subsequent , girth welds* sections only) -examinations on a ten-year interval.

* Under MRP-227, Revision 1 [2], this component would be the Core Support Barrel Assembly Middle Girth Weld (MGW) with expansions to the middle axial weld (MAW) and lower axial weld (LAW). Per the responses to NRC RAis on MRP-227, Revision 1 [15], the core support columns could also become an expansion to the MGW. Once MRP-227, Revision 1 has received a safety evaluation with acceptance of these changes, revisions to the naming of the Primary and Expansion components for the MRP-227-A "lower cylinder girth welds" provided in the approved version ofMRP-227, Revision 1 should be substituted here.



July 9, 2018

Table 5-12: Expected Revised Entries for Table 5-2 for MRP-227, Revision 2 - CE Plants Examination Acceptance and Expansion Criteria (NEI 03-08 Needed)

Current Table Entrv (MRP-227-A Table 5-2) Primary Item Applicability Examination Expansion Link(s) Expansion Criteria Expansion Item

Acceptance Examination

Criteria Acceptance Criteria

Core Support All plants Visual (EVT-1) Lower cylinder axial Confirmation that a surface-breaking indication >2 inches The specific relevant

Barrel examination. welds in length has been detected ancl sized in the lower cylinder condition for the

Assembly The specific girth weld shall require an EVT-1 examination of all expansion lower cylinder

Lower cylinder relevant accessible lower cylinder axial welds by the completion of axial welds is a

girth welds condition is a the next refueling outage. detectable crack-like

detectable surface indication.

crack-like surface -

indication.

R . d Ti bl E t evzse a e nrrv Primary Item App Ii ca bility Examination Expansion Link(s) Expansion Criteria Expansion Item

Acceptance Examination

Criteria Acceptance Criteria

Core Support All plants Visual (EVT-1) a. Lower cylinder a. The confirmed detection and sizing of a surface-breaking a. The specific relevant

Barrel examination. axial welds indication >2 inches in length in a lower cylinder girth weld condition for the

Assembly . The specific b. Fuel alignment shall require an EVT-1 examination of all accessible lower expansion lower cylinder

Lower cylinder relevant plate (Plant designs cylinder axial welds by the completion of the next refueling axial welds is a

girth welds* condition is a with core shrouds outage. detectable crack-like

detectable assembled in two b. (Applicable only after entering SLR period) The surface indication.

crack-like vertical sections confirmed detection and sizing of a surface-breaking b. The specific relevant

surface only) indication >2 inches in length in a lower cylinder girth weld condition is a detectable

indication. shall require an EVT-1 of the fuel alignment _plate within crack-like surface

the next three refueling outages. indication.

* Under MRP-227, Revision 1, this component would be the Core Support Barrel Assembly Middle Girth Weld (MGW) with expansions to the middle axial weld

(MAW) and lower axial weld (LAW). Per the responses to NRC RAis on MRP-227, Revision 1 [15], the core support columns could also become an expansion to the

MGW. Once MRP-227, Revision 1 has received a safety evaluation with acceptance of these changes, revisions to the naming of the Primary and Expansion

components for the MRP-227-A "lower cylinder girth welds" provided in the approved version ofMRP-227, Revision 1 should be substituted here.


5.2 Impacts on Westinghouse Inspection Requirements



July 9, 2018

5.2.1 Differences between MRP-227-A and MRP-227, Revision 1 Inspection Requirements

This section addresses the differences between the baseline MRP-227-A I&E requirements and the revised MRP-227, Revision 1 requirements. The following list of changes to the MRP-227 inspection program tables were a mix of more conservative criteria, less conservative criteria, and clarifications:

• Changes for Primary components o Examination requirements for the CRGT Assembly Guide Plates (Cards) must be

consistenfwithMRP-2014-006 [39] and WCAP-1'7451-P [36] . o CRGT lower flange welds

• Added expansion component for the remaining CRGT lower flange welds • No longer expands to the lower support column bodies (cast), the UCP, or the

lower support forging/casting • Specified that the weld inspection includes 0.25 inches of adjacent base metal

o Upper core barrel flange weld • Renamed to upper flange weld (UFW) • No longer expands to the lower support column bodies (non-cast) or the core

barrel outlet nozzle welds • Added expansion to low fluence girth and axial welds (UGW, LFW, and UAW) • Added expansion to lower support forging or casting

o Removed component upper and lower core barrel cylinder girth welds and replaced with the UFW and LGW primary components and the UGW expansion component

o LGW component added to list of primary compon_ents with the following expansion items:

• UCP • Lower support column bodies ( cast and non-cast) • MAW • LAW

o Upper core barrel girth weld and lower core barrel flange weld component moved.to expansion category

o Coverage of primary core barrel welds reduced to 25% sampling o Additional examination coverage details provided _for the Baffle-former assembly

• Changes for Expansion components o Remaining CRGT lower flange welds added as an e'xpansion component from the CRGT

lower flange welds o As noted in the primary list, several components received different Primary items

• UCP • Lower support forging or casting • Lower support column bodies

o Coverage of lower support forging or casting reduced to 25% of the bottom surface using a VT-3 inspection

o Core barrel outlet nozzles removed from the expansion components _ o Upper and lower core barrel axial welds divided out into individual welds

• MAW and LAW expand from the LGW


• UAW expands from the UFW


Page 44 of70 LTR-AMLR-18-4, Rev.·1

July 9, 2018

o UGW, UAW, LFW, and lower support forging/casting added as expansions from UFW

o Coverage of UCP reduced to 25% of core side surfaces using VT-3

o Coverage of lower support column bodies ( cast and non-cast) reduced to 25% of column

assemblies as visible using a VT-3 examination from above the lower core plate

• Changes fo~ Existing components o Added the clevis bearing stellite wear surface for inspection

For the gap analysis with MRP-227-A, only the more conservative criteria will 1:,,e relevant to an SLR

application. The more conservative criteria from the above list are:

• Examination requirements for the CRGT Assembly Guide Plates (Cards) must be consistent with

MRP-2014-006 [39] and WCAP-17451-P [36]

• Added expansion component for the remaining CRGT lower flange welds

• CRGT lower flange weld inspection specifies that the weld inspection includes 0.25 inches of

adjacent base metal • Added the clevis bearing stellite wear surface for Existing inspection

Adding these items from MRP-227, Revision 1 to the SLR aging management program is the first step in

meeting the requirements of the SLR gap analysis.

For the remaining changes, including the clarifications should be considered for the improved

understanding that they provide. The addition of the less conservative criteria will require inclusion of an

adequate technical basis. Much of this technical basis has already been developed through the MRP-227,

Revision 1 basis documents and responses to NRC RAis, but use of the less conservative criteria carries

an inherent risk since MRP-227, Revision 1 has not yet received an NRC safety evaluation.

5.2.2 Changes to the current Primary, Expansion, or Existing Inspection Requirements

Seven Westinghouse components or component groups that are currently Primary, Expansion, or Existing

Inspection items in MRP-227-A and MRP-227, Revision 1 [2] are expected to require changes for MRP-

227, Revision 2:

• Changes due to SLR expert panel review o Baffle-edge bolts (clarification on bracket bolts)

o Baffle-former bolts ( clarification on _comer bolts)

• Potential changes due to NRC RAis on MRP-227, Revision 1 or due to interim guidance

o CRGT assembly guide cards o Upper core plate o Baffle-former bolts o Baffle-form~r bolt expansion criteria (barrel-former bolts) o Core barrel welds o Lower support columns


Baffle-Edge Bolts

· Enclosure to MRP 2018-022

Page 45 of70 LTR-AMLR-18-4,Rev. 1

July 9, 2018

The SLR expert panel review added the bracket bolts component under the baffle and former assembly as shown in Table 4-8. These bolts were not missed during the development ofMRP-191, Revision 1 [5], MRP-227-A [1], or MRP-227, Revision 1 [2]. In those documents, the bracket bolts were assumed to be included under the baffle-edge bolts, since they were located in similar locations and performed similar functions. Figure 5-1 shows where the bracket bolts are located in a typical baffle-former ~sembly. Note that they are only present in 4-loop plants originally designed for downflow operation. For the development ofMRP-227, Revision 2, the SLR expert panel decided that additional clarity should be added by including the bracket bolts by name.

All of the screening, categorization, and ranking conclusions for the baffle-edge bolts are also applicable to the bracket bolts. For MRP-227, Revision 2, the entry for the baffle-edge bolts will be updated to name the bracket bolts as well.

The revised Primary inspection item entry for the baffle-edge bolts, includihg the bracket bolts, for MRP-227, Revision 2, Table 4-3 is shown in Table 5-13, along with the previous MRP-227-A entry for comparison. This change is applicable to the first PEO. The guidance in this table constitutes an NEI 03-08 "Needed" requirement.

BAFF1.E TO FORMER BOLT(LONO & SHORT)

Bracket Bolt

Figure 5-1: Bolt locations in a Typical Westinghouse Baffle-Former-Barrel Assembly Containing Bracket Bolts


Baffle-Former Bolts



July 9, 2018

The SLR expert panel review added the corner bolts component under the baffle and former assembly as

shown in Table 4-8. These bolts were not missed during the development ofMRP-191, Revision 1 [5],

MRP-227-A [l], or.MRP-227, Revision 1 [2]. In those documents, the corner bolts were assumed to be

included under the baffle-former bolts, since they were located in similar locations and performed similar

functions. For the development ofMRP-227, Revision 2, the SLR expert panel decided that additional

clarity should be added by including the corner bolts by name. Corner bolts are only applicable to a

subset of 3-loop downflow plants that include a corner angle baffle plate. The locations of the corner

bolts and this corner angle baffle plate are shown in Figure 5-2.

All of the screening, categorization, and ranking conclusions for the baffle-former bolts are also

applicable to the corner bolts. For MRP-227, Revision 2, the entry for the baffle-former bolts will be

updated to name the corner bolts as well.

The revised Primary inspection item entry for the baffle-former bolts, including the corner bolts, for

MRP-227, Revision 2, Table 4-3 is shown in Table 5-13, along with the previous MRP-227-A entry for

comparison. This change is applicable to the first PEO. The guidance in this table constitut~s an NEI 03-

08 "Needed" requirement.

___ Comer Angle

Baffle Plate

Figure 5-2: Bolt locations in a Typical Westinghouse Baffle-Former-Barrel Assembly Containing

Corner Bolts



July 9, 2018

Potential Changes due to NRC RA!s on MRP-227. Revision 1 or due to Interim Guidance I

The details of t!ie potential revisions to tJ:ie inspection requirements for these five components are provided in greater detail in the RAI response letters [14] [15] [16] and in the baffle-former bolt interim guidance letters [10] [11]. The RAI response letters include the NRC RAls, the detailed responses, and proposed revised entries for the MRP-227, Revision I tables. The baffle-former bolt interim guidance letters provide required revisions to the l&E guidance for the bolts and for the expansion criteria when large clusters of degraded bolts are observed. A summary ofthese changes is provided here:'

I

• CRGT Assembly Guide Cards o More aggressive than expected wear has recently been observed for plants using control

rods with an ion nitride surface treatment. This accelerated degradation requires changes to the inspection guidance ofWCAP-17451-P [36] as detailed and approved as NEI 03-08 "Needed" requirements in letter OG-18-46 [12] and transmitted to the NRC in letter OG-18-76 [13].

• Upper core plate o NRC RAI 14 on MRP-227, Revision 1 has questioned the reduction in inspection

coverage and inspection technique from MRP-227-A [1] to MRP-227, Revision 1 for the UCP. This could result in changes to the requirements ofMRP-227, Revision 1 in an NRC-approved version.

• Baffle-former bolts o As a response to the operating experience with baffle-former bolts, the industry

developed interim.guidance that modified the initial inspection timing, evaluation criteria, and re-inspection interval for the bolts [10]. The NRC staff assessed the technical basis behind this interim guidance and accepted the guidance for aging management of baffleformer bolts in [37]. NRC RAI 8 also asked a question about how the industry plans to respond to the baffle-former bolt experience, and the response was that the baffle-former bolt interim guidance would be used [14].

• Baffle-former bolt expansion criteria o Interim guidance on the baffle-former bolt expansion criteria was published to address

cases where large clusters of degraded bolts or bolts with relevant indications were observed [ 11]. This interim guidance defined what is considered a large cluster of degraded bolts and provided a revised Expansion inspection requirement to visually inspect at least a portion of the barrel-former bolts if a large cluster of degraded baffleformer bolts is observed.

• Core barrel welds (RAI 5 and 26) o The response to RAI s,provided a basis forthe reduced inspection coverage of the core

barrel welds, and RAI 26 provided a basis for the assignment of welds to the Primary or Expansion component lists. These changes were implemented in MRP-227, Revision 1 but may not be accepted by the NRC staff. The recent operating experience of a CEdesigned plant detecting crack-like indications in the core support barrel will likely have an impact on the resolution of these two RAis. This interim guidance does not create revised requirements for CE core support barrel welds, since the industry is currently evaluating the need for further guidance on core barrel aging management.


• Lower support columns (RAI 9)



July 9, 2018

o The response to RAI 9 provided a basis for the reduced inspection coverage and

requirements for the lower support columns. These changes were implemented in MRP-

227, Revision 1 but may not be accepted by the NRC staff.

5.2.3 Additional Components to be added to the Primary, Expansion, or Existing Categories

It is expected that five components or component groups wiU be added to or elevated within the

Westinghouse Primary, Expansion, and Existing component categories:

• Primary o Clevis insert bolts and dowels ( elevated from Existing)

p Thermal sleeves o Core barrel assembly radial support keys

o Clevis inserts (Elevated from Existing)

• Expansion ·o None

• Existing o. Malcomized fuel alignment pins

Clevis Insert Bolts and Dowels

The experience in Westinghouse plants with clevis insert bolt cracking and clevis insert dowel

degradation [3~] indicates that the clevis insert bolts and dowels need to be managed for potential aging

degradation. As noted in Section 3, the cracking in the bolts was due to the susceptibility of the Alloy

X-750 bolts to primary water SCC. The degradation of the dowels was likely related to the same aging

degradation mechanism. Though there are variations in the clevis insert designs from plant to plant in the

Westinghouse fleet, the clevis insert bolts at all plants were fabricated from susceptible Alloy X-750

material and are subject to this failure mode. Based on the operating experience to date and the high

susceptibility of this material, the SLR expert panel assigned the clevis insert bolts to safety risk Category

Band economic risk Category C [6]. The SLR expert panel also added the clevis insert dowels as a

separate line item in the MRP-191 table and, based on the operating expertence, assigned the dowels to

· safety risk Category B [ 6].

The clevis insert bolts were included in both MRP-227-A and MRP-227, Revision 1 as an Existing

inspection item. Due to the fact that degradation of these bolts has already been observed at multiple

plants, the bolts should be elevated from Existing to Primary for MRP-227, Revision 2. This elevation is

also supported by the fact that no other componentin the Westinghouse reactor vessel internals could

effectively serve as a leading indicator for the primary water SCC that the clevis insert bolts are expected

to experience. Since degradation has already been observed in the clevis insert bolts prior to the first

PEO, it is recommended that the initial inspection be set for the start of the first PEO at 40 years.

Consistent with other components, the re-inspection interval would be 10 years.

The general visual (VT-3) ASME Code Section XI inspection that was specified for the clevis insert bolts

may detect degradation in the bolts, but if the degradation is hidden beneath the head, it may not. Past



July 9, 2018

inspection experience has shown that once bolts are fully separated and the heads have begun to wear on locking devices or the clevis insert, the VT-3 examination can detect degradation. Evaluations of clevis insert bolt cracking degradation were documented in PWROG-15034-P [38]. This study determined that degraded clevis insert bolts would not result in a loss of function of the lower radial support system and thus should not pose a safety concern. Several physical constraints were shown to prevent this loss of function during operation. This technical basis supported the conclusion that the current general visual (VT-3) examinations from MRP-227 are suf~cient to manage the safety function of the lower radial support system. Pro-active replacement of these bolts would also be an acceptable approach to managing the potential aging degradation.

The clevis insert dowels were not included in MRP-227-A or MRP-227, Revision 1 as a Primary, Expansion, or Existing inspection component. Since degradation has already been observed in the dowels, they should also be elevated for MRP-227, Revision 2. The dowels are located directly adjacent to the clevis insert bolts, and the expected degradation (fractured tack welds on the dowels or rotation of the dowels) can be detected by the same general visual (VT-3) inspection specified for the bolts. Thus, these two components can be combined into the same Primary inspection requirement.

Typically, the safety risk Category B assignment of the clevis insert bolts and dowels would not automatically result in the bolts being assigned a Primary inspection. However, the multiple instances of active degradation and the potential for varying Existing component inspection requirements from plantto-plant merit a more conservative aging management approach.

The degradation mechanism is equally likely to occur at any clevis insert bolt or dowel location, so the required coverage should be 100% of the accessible clevis insert bolts and dowels. All of the bolts and dowels are expected to be accessible.

The revised inspection requirements adding the clevis insert bolts and dowels to MRP-227, Revision 2 Table 4-3 after the first PEO are shown in Table 5-14. The guidance in this table cpnstitutes an NEI 03-08 "Needed" requirement.

Thermal Sleeves

Wear of the reactor vessel head adapter thermal sleeve has been observed at several Westinghousedesigned plants [29]. This has reached unacceptable levels in some cases and resulted in thermal sleeves breaking free from their normal position. The wear occurs in three locations: the thermal sleeve flange, the outer diameter of the sleeve, and the inner diameter of the sleeve. Details on the operating experience, inspection timing, recommended inspection techniques, and calculation of re-inspection intervals have been provided in the technical bulletin [29] and throu¥11 the Pressurized Water Reactor Owners Group [40] [41].

Based on this operating experience and the potential for significant economic consequences, the SLR expert panel review concluded that these thermal sleeves should be assigned to economic risk category C. At the time of the first SLR expert panel, the safety consequence was thought to be low, which resulted in a safety risk category of B being assigned. However, recent operating experience indicates that wear of the thermal sleeves can impede the ability to insert a control rod assembly [30]. The expert panel was,



July 9, 2018

reconvened to evaluate the thermal sleeves in light of this new operating experience and concluded that

both the safety risk and economic risk categories for the thermal sleeves should be Category C [6].

It is recognized that there is some inconsistency in how individual plants categorize the thermal sleeves.

The SLR expert panel categorized the sleeves as reactor vessel internals. This is based on past

documentation and treatment of the sleeves by Westinghouse. The thermal sleev~s show up on the

reactor general assembly drawings of plants, at the same level as internals components such as the upper

internals assembly, hold down spring, core barrel assembly, and clevis insert assembly. They are also

listed in the scope of supply for the generic reactor internals equipment specifications and plant-specific ·

addenda under "Associated Hardware." There are examples of plants that include the thermal sleeves

under the reactor vessel internals program of the plant Final Safety Analysis Report. Confusion a:bout the

thermal sleeves being part of the reactor vessel internals may stem from the fact that the sleeves were

included in reactor vessel head replacements, rather than moving the old thermal sleeves to the new head.

No other component in the reactor vessel internals can provide a leading indication for the wear

degradation that has been observed in the thermal sleeves, and the wear to date has already resulted in

failures. Based on these reasons and the potential impacts on safety, the thermal sleeves should be

assigned to the Primary inspection components for MRP-227, Revision 2. Based on the fact that

unacceptable degradation has already been observed prior to the first PEO, the initial. inspection should be

required according to the initial inspection timing provided in the reference documents [29] [ 40] [ 41].

The required initial inspections vary depending on the plant design and the modifications that have been

implemented, so this initial inspection timing must be determined on a plant-specific basis. Subsequent

inspections are dependent on the amount of wear measured and the projected amount of future wear [40]

[41].

Some aspects of the wear degradation in the thermal sleeves can be detected with a general visual

examination, but much of the wear occurs in locations that are inaccessible to a visual inspection or in a

manner which obscures effective visual detection of the wear. Therefore, the following techniques are

recommended for detecting wear in the thermal sleeves:

• Ultrasonic test (UT) to detect inner diameter or outer diameter wear of the sleeves

• Measurement of the height of the thermal sleeve guide funnels relative to the reactor vessel

closure head to detect wear in the thermal sleeve flange.

The inspection should be conducted per the plant design-specific inspection recommendations of TB-07-2

[29]. These inspection recommendations include the type, coverage, and timing of the inspection. Not all

Westinghouse-designed plants have thermal sleeves, so this inspection requirement is only applicable to

those plants with thermal sleeves.

The revised inspection requirements adding the reactor vessel head adapter thermal sleeves to MRP-227,

Revision 2 Table 4-3 are shown in Table 5-14. This change is applicable to the first PEO. The guidance

in this table constitutes an NEI 03-08 "Needed" requirement.


Wear Surfaces on the Radial Support Keys and Clevis Inserts



July 9, 2018

Wear on the stellite hardfacing surfaces of the radial support keys and clevis inserts is a concern for the Westinghouse-design plants [33]. These are mating components that provide alignment for the core barrel. This wear could particularly become an issue as plants continue operation into the second PEO. The -SLR expert panel determined that the wear of the clevis inserts and radial support keys should be elevated to Category C for both safety and economics. The elevated safety concern was due to the fact that these provide a core support and safe. shutdown function by lir.niting the amount of circumferential or radial displacement in the core barrel. The elevated economic concern stemmed from the high difficulty of repairing these components due to the precision fits that were required during original fabrication. The same reasoning was deemed applicable to both components.

The clevis inserts were not included in MRP-227-A but are included in MRP-227, Revision 1 [2] as an Existing inspection component (see Section 5.2.1). Based on the results of the expert panel review, the clevis inserts should be elevated to be a Primary inspection component. The radial support keys were not assigned to an inspection category in either MRP-227-A or MRP-227, Revision 1 and should also be added to the Primary inspection category, based on the expert panel review results. Operating experience has shown that significant wear is already occurring on these components at some plants, and logically, this wear will continue to increase as the plants operate into the second PEO.

The timing for the first inspection on the radial support keys and clevis insert wear surfaces would be at the start of the first PEO. This is based on the fact that some plants have already observed significant wear on the mating surfaces of these components. Consistent with other components, 'the re-inspection interval would be 10 years.

All locations are potentially susceptible to wear on these components, so the coverage requirement should be 100% of the radial support keys and 100% of the clevis inserts. The inspection must focus on the wear surfaces and look for evidence of excessive wear. The inspection type will be general visual (VT-3). Note that this inspection is intended to detect the presence of wear on these surfaces but is not expected to be effective at measuring the full extent of material loss that may have occurred. Such measurements are beyond the scope of this existing inspection but may be required for evaluation and disposition of a relevant condition.

The new Primary inspection requirements adding the radial support key wear surfaces and the clevis insert wear surfaces to MRP-227, Revision 2 Table 4-3 are shown in Table 5-14. The guidance in this

I • ' table constitutes an NEI 03-08 "Needed" requirement.

Malcomized Fuel Alignment Pins

Per technical bulletin TB-16-4 [34], accelerated loss of material degradation has been observed on fuel alignment pins that have a malcomized surface. This was a surface treatment applied to the fuel alignment pins at many early plants to increase the hardness and resistance to wear. The degradation appears as a thin layer of material flaking off of the surface. The fuel alignment pins on both the UCP and the lower core plate can be affected by this mechanism, if they are malcomized. Based on the known operating experience, the SLR expert panel assigned these fuel alignment pins to a high likelihood of



July 9, 2018

degradation, but maintained a low safety consequence because of the evaluations documented in TB-16-4.

These showed that the loss of the malcomized surface layer and the resulting larger gap between the fuel

assemblies and the fuel alignment pins should have a small effect on fuel mechanical and reload design

criteria and on toss of coolant accident analyses. Thus, the expert panel assigned the malcomized fuel

alignment pins to safety Category B [ 6].

The fuel alignment pins are inspected regularly under existing ASME Section XI code requirements. This

inspection is a general visual (VT-3) and is appropriate for detection of the loss of material that may

occur. This inspection could be considered for elevation to Primary due to the degradation already

observed, similar to the approach taken for the clevis insert bolts. However, the degradation of the

malcomized coatings have had limited impact and the fuel alignment pins with this degradation are still

present and capable of performing their function. Thus, adding the malcomized fuel alignment pins as an

Existing inspection is deemed appropriate. As noted above, the general visual (VT-3) inspection already

required for the pins is appropriate. Coverage should be 100% of the accessible fuel alignment pins, since

the degradation can occur on any of the malcomized locations. Note that no other component in the

Westinghouse reactor vessel internals could effectively serve as a leading indicator for this degradation

mechanism. Consistent with other components, the re-inspection interval would be 10 years.

The revised inspection requirements adding the malcomized fuel alignment pins (both UCP and lower

core plate fuel alignment pins) to MRP-227, Revision 2 Table 4-9 after the first PEO are shown in Table

5-15. The guidance in this table constitutes an NEI 03-08 "Needed" requirement.

5.2.4 Components to be removed from the MRP-227 Requirements

No components are expected to be removed from the MRP-227 I&E guidelines for SLR based on the

results of the MRE-191 SLR expert panel review.

5.2.5 Elevated Risk Category Items that are not Expected to Change

Multiple components that received a safety or economic Category C risk ranking in either MRP-191,

Revision 1 [5] or during the SLR expert panel review are not expected to be changed for MRP-227,

Revision 2.

• CRGT assembly flexures

o The CRGT assembly flexures are only applicable to one plant known to the SLR expert

panel. Thus, they are considered a plant-specific item that will not be it?-cluded as an

inspection component in MRP-227, Revision 2.

• CRGT guide plates/cards

o The CRGT guide plates/cards are currently a Primary inspection component for wear in

MRP-227, Revision 1 [2]. Wear continued to be the degradation mechanism of concern

for these components in the SLR expert panel review. Additionally, the guide

plates/cards are being managed through the guidance provided in WCAP-1-7451 [36], as

communicated to the NRC in response to RAI 19 on MRP-227, Revision 1 [14]. The

guide plates/cards will remain as Primary inspection components for MRP-227, Rev. 2.


• CRGT sheaths and c-tubes


' Page 53 of70 LTR-AMLR-18-4, Rev. 1

July 9, 2018

o The CRGT sheaths and c-tubes were both placed in risk category C for MRP-191, Rev}sion 1. This elevated category was maintained for the sheaths in the results of the SLR expert panel review, where it was placed in Category C for safety and economics. The c-tubes were reduced to safety and economic Category B because the wi,!ar that has been observed on the c-tubes was not limiting when compared to that observed on the guide plates/cards and sheaths. For MRP-227, Revision: 1 [2], the c-tubes and sheaths were not placed in an inspection category because the CRGT guide plates/cards were limiting. That is still the case for MRP-227, Revision 2.

• Thermal shield flexures . o The thermal shield flexures will continue to be a Primary inspection component in MRP-

227, Revision 2. • Internals hold-down spring

· o The internals hold-down spring (Type 304 SS only) will continue to be a Primary inspection component in MRP-227, Revision 2.

• X-750 guide tube support pins (RAJ 15) ·, o The X-750 guide tube support pins have been addressed on a plant-specific basis across

the industry. They will remain as a plant-specific program for MRP-227, Revision 2. This is supported by the response to NRC RAJ 15 on MRP-227, Revision 1 [14].

• Flux thimble tubes o The flux thimble tubes will continue to be an Existing inspection component in MRP-

227, Revision 2.

There were also a number of components which were elevated to safety risk category B by the SLR expert panel, whi.ch do not require additional aging management requirements for SLR:

• CRGT flanges, lower (wrought) o Under MRP-191, Revision 1, only the cast flanges were elevated to safety Category B.

The SLR expert panel kept the cast flanges at B but increased the wrought flanges to match. The current Primary inspection item for CRGT lower flange welds already covers both of these components, so no change is required.

• Conduit seal assembly: body, tubesheets, tubesheet welds and tubes o These were screened out for all aging degradation mechanisms in MRP-191, Revision 1

[5], and fatigue and SCC were added in the SLR expert panel. These have experienced degradation in plants, but the cause could be plant-specific rather than generic. Additionally, the safety consequence is driven by potential leakage, which is monitored on a regular basis.

• Upper core plate insert o These inserts serve as a wear surface for alignment of the upper internals. The expert

panel review concluded that the potential wear on the clevis inserts and radial support keys is expected to lead the wear on the UCP insert, so no additional measures are required.


I



) July 9, 2018

• Protective skirt bolts, lower core plate and manway bolts, and thermal shield bolts

o These bolts are subject to potential IASCC, SCC, and/or fatigue. The baffle-former bolts

and their expansion components are already managed for these aging degradation

mechanisms and are expected to be reasonable lead items for these other bolted locations.

• Baffle bolting locking devices o These locking devices were elevated from Category A to safety Category B because of

the actual degradation that has been observed at multiple plants. However, the locking

devices are £llready managed directly under the baffle-former assembly and baffle-edge

bolt inspections and indirectly under the baffle-former bolt inspection.

• BMI column bolts o As with the other bolts affected by IASCC, SCC, and/or fatigue discussed above, the

BMI column bolts are expected to be adequately managed by the current Primary and

Expansion inspections of bolts. In addition, the Expansion inspection requirements for

the BMI columns also provide a level of management for the BMI column bolts.

• Safety injection nozzle interface o The expert panel concluded that this nozzle is similar to the core barrel outlet nozzle in

many ways. The core barrel outlet nozzles were originally an Expansion inspection item

in MRP-227-A [1] but were eventually removed in MRP-227, Revision 1 [2].



July 9, 2018

Table 5-13: Expected Revised Entries for Table 4-3 for MRP-227, Revision 2 - Westinghouse Primary Components for Implementation for. First PEO (NEI 03-08 Needed)

Current Table Entries (MRP-227-A Table 4-3 Primary Item App Ii ca bility Effect (Mechanism) Expansion Link Examination Method/

Frequency Baffle-Former All plants with Cracking (IASCC, Fatigue) that None Visual (VT-3) Assembly baffle-edge bolts results in examination, with baseline Baffle-edge _bolts • Lost or broken locking devices examination between 20

• Failed or missing bolts and 40 EFPY and • Pro_trusion of bolt heads subsequent examinations Aging Management (IE and on a ten-year interval. ISR)

Baffle-Former All plants Cracking (IASCC, Fatigue) Lower support Baseline volumetric (UT) Assembly Aging Management (IE and -- column bolts, examination between 25 Baffle-former bolts ISR) Barrel-former and 35 EFPY,. with

bolts subsequent examination on a ten~year interval.

Revised Table Entries Primary Item App Ii ca bility Effect (Mechanism) Expansion Link Examination Method/ . ...,

Frequency Baffle-Former All plants with Cracking (IASCC, Fatigue) thaf None Visual (VT-3) Assembly baffle-edge bolts or results in examination, with baseline Baffle-edge bolts bracket bolts • Lost or broken locking devices examination between 20 Bracket bolts1

• Failed or mi_ssing bolts and 40 EFPY and • Protrusion of bolt heads subsequent examinations Aging Management (IE and on a ten-year interval. ISR)

Baffle-Former All plants ( comer Cracking (IASCC, Fatigue) Lower support Examination according to Assembly bolts are only ,Aging Management (IE and column bolts, the requirements ofMRP Baffle-former bolts applicable to some ISR) Barrel-former 2017-009 [101 ,

Comer bolts2 plants) -- bolts I. Bracket bolts are only applicable to 4-loop plants originally designed for downflow configuration. 2. Corner bolts are only applicable to the 3-loop design plants which include corner angle baffle plates

Examination Coverage

Bolts and locking devices on high fluence s6;ms. 100% of components accessible from core side.

100% of accessible bolts. Heads accessible from the core side. UT accessibility may be affected by complexity of head arid locking device design


Bolts and locking devices on high fluence seams. 100% of components accessible from core side.

100% of accessible bolts.

--,



July 9, 2018

Table 5-14: Expected New Entries for Table 4-3 for MRP-227, Revision 2 - Westinghouse Primary Components (NEI 03-08 Needed)

Primary Item App Ii ca bility ·Effect (Mechanism) Expansion Link Examination Method/ Frequency Examination Coverage

Alignment and All plants Cracking (SCC), Loss of None Visual (VT-3) examination no later than 2 All clevis insert

Interfacing material (Wear) refueling outages from .the beginning of bolts and clevis

Components the first license renewal period. insert dowels

Clevis insert bolts Subsequent examinations on a ten-year

Clevis insert dowels interval.

Alignment and All plants with Loss of material (Wear) None Volumetric (UT) examination according Thermal sleeve wear

Interfacing thermal sleeves to the requirements and initial inspection surfaces according

Components timing ofTB-07-02 [29] and WCAP- to the requirements

Thermal sleeves 16911-P [40] of[29] [40] [41] Measurement of thermal sleeve guide funnels height according to the requirements and timing ofTB-07-02 [29] and PWROG-16003-P [41] Subsequent examinations based on calculated wear projections.

Radial Support Keys All plants Loss of Material (wear) None Visual (VT-3) examination no later than 2 Wear surfaces on all

Radial support keys refueling outages from the beginning of radial support keys the first license renewal period. Subsequent examinations on a ten-year interval.

Alignment and All plants Loss ofMaterial (wear) None Visual (VT-3) examination no later than 2 Wear surfaces on all

Interfacing refueling outages from the beginning of clevis inserts

Components the first license renewal period.

Clevis bearing Stellite Subsequent examinations on a ten-year

wear surface interval.



July 9, 2018

Table 5-15: Expected New Entries for Table 4-9 for MRP-227, Revision 2 - Westinghouse Existing Components (NEI 03-08 Needed)

Item Applicability Effect (Mechanism) Reference Examination Method Examination Coverage >

UCP and Fuel All plants with Loss ofMaterial (wear) ASME Code Section XI Visual (VT-3) All accessible surfaces at Alignment Pins malcomized fuel examination specified frequency Fuel alignment pins

alignment pins on the UCP (See TB-16-4 [34])

LCP and Fuel All plants with Loss of Material (wear) ASME Code Section XI Visual (VT-3) All accessible surfaces at Alignment Pins malcomized fuel examination specified frequency Fuel alignment pins

alignment pins on the LCP (See TB-16-4 [34])

/



July 9, 2018

6 Asset Management Recommendations for License Renewal

· The separatiort of the consequences of degradation into safety and economi~ categories allowed the expert

panel to evaluate the two types ofrisk separately. Safety risk is the main driver for the reactor vessel

internals aging management l&E guidelines, but utilities have a strong interest in managing the costs of

the degradation that will occur with long-term operation. The changes and additions to MRP-227

provided in Section 5 were based primarily on the safety risks of component degradation. These changes

and additions did not capture all of the components that were assigned to economic risk category C. The

current section focuses on addressing the potential commercial risks associated with the economic

Category C components and component assemblies.

The approach in this section started with the list of economic Category C items designated by the SLR

expert panel review. These are.included in Table 4-7 for CE components and in Table 4-8 for

Westinghouse components and are repeated in Table 6-1 (CE components) and Table 6-2 (Westinghouse

components) for evaluation of asset management. Asset management was divided into an inspection

piece and a supplemental contingency piece in these tables. In Table 6-1 and Table 6-2, the components

that already have adequate inspection requirements through MRP-227, the interim guidance detailed in

Section 5, or some other existing program were noted in the list. This left only a handful of components

that were either subject to Expansion inspections or did not have any MRP-227 requirements, which the

SLR expert panel considered to have the highest combination of economic consequence and likelihood of

experiencing degradation.

• CE o Fuel alignment plate for all CE plants except for System 80 designs

o Core support barrel assembly: .

• Upper cylinder axial welds (UAW)

• Lower cylinder axial welds (MAW and LAW)

• Westinghouse · o Upper core plate

o Brackets, clamps, terminal blocks, and conduit straps

o Barrel-former bolts

o Core barrel assembly:

• Upper core barrel axial welds (UAW)

• Lower core barrel axial welds (MAW and LAW)

Multiple approaches can be used to conduct asset management on these components. If inspection is the

preferred method of management, most of these can be adequately evaluated through a general visual

(VT-3) inspection. Recommendations for the periodic visual inspection of these components are provided

in Table 6-3 for CE plants and Table 6-4 for Westinghouse plants. These tables are considered NEI 03-08

"Good Practice" recommendations.

The recommendations in Table 6-3 and Table 6-4 are focused on asset management through inspection,

but this only fulfills the early detection part of managing the potential for high costs from aging

degradation of reactor vessel internals and only provides one method of achieving early detection. To

effectively manage costs, a plant owner should also consider the need to prepare contingency inspections



July 9, 2018

or repairs and the benefits of supplemental monitoring approaches, such as neutron noise for some components or monitoring the integrity of the upper instrumentation measurement channels for the brackets, clamps, terminal blocks, and conduit straps. Approaches for developing contingency strategies and altymative monitoring approaches are summarized in multiple industry documents. Table 6-5 is referenced in both Table 6-1 and Table 6-2 and provides some high-level links to other documents that should be used when developing an overall asset management program for the reactor vessel internals.

T bl 6 1 C a e - . Assembly/

Subassembly

Upper Internals

Assembly

Lower Support

Structure

CSB Assembly

Core Shroud

Assembly


om b t' US IOU E nemeermg E ' R' kC t conom1c IS a egory cc omponen tD' 1spos1 100mg Screened-in

Safety Economic Current Asset Management Component Material Degradation

Category Category Status Mechanisms

System 80 plants: Primary

inspection component starting at Weld, IASCC,

first PEO Fuel alignment plate 304 ss Wear,_Fatigue, B C

All other CE plants: Expansion IE

component starting at second

PEO

System 80 Core Weld, IASCC, Primary inspection component

support deep beams 304 ss

Fatigue, IE, VS C C

starting at first PEO

Upper cylinder girth Weld, IASCC, Primary inspection component

welds (UFW and 304 ss B C

UGW) Fatigue, IE starting at first PEO

Upper cylinder axial Weld, IASCC, Expansion component starting at

welds (UAW) 304 ss

Fatigue, IE B C

first PEO

Lower cylinder girth Weld, IASCC, Primary inspection component

welds (MGW and 304 ss B C

LGW/LFW) Fatigue, IE starting at first PEO

Lower cylinder axial Weld, IASCC, Expansion component starting at

welds (MAW and 304 ss B C

LAW) Fatigue, IE first PEO

Lower core barrel 304 ss Weld, Fatigue B C

Primary inspection component

flange flexure weld starting at first PEO

Core stabilizing lug X-750 SCC, Wear B C


shim bolts starting at first PEO

IASCC, Wear, Significant portions inaccessible:

Core shroud tie rods 348 ss Fatigue (I), IE, B C Primary inspection component

VS, ISR/IC starting at first PEO

Core shroud tie rod IASCC, Wear,

Primary inspection component 316 ss Fatigue (I), IE, B C

nuts VS,ISR/IC

starting at first PEO



July 9, 2018

Asset Management Recommendation

System 80 plants: - Current inspection adequate for asset management

- Evaluate need for contingency options (See Table 6-5)

All other CE plants: - Consider performing proactive inspection prior to

expansion criteria are triggered


- Current inspection adequate for asset management


- Current inspection adequate for asset management*


- Consider performing proactive inspection prior to

expansion criteria.are triggered*




- Consider performing proactive inspection prior to

expansion criteria are triggered*



- Evaluate need for contingency optio~s (See Table 6-5)







* Core support barrel weld mspect10n reqmrements may be subJect to change based on.response to 2018 Sprmg outage operating experience and MRP-227, Rev. 1 safety evaluation


Table 6-2: Westinghouse Economic Risk Category C Component Dispositioning

Screened-in Safety Economic Current Asset Management Assembly Subassembly Component Material Degradation

Mechanisms Category Category Status

Flexures X-750 SCC, Fatigue C C Only applicable to one plant Primary inspection

Guide plates/cards 304 ss Weld, Wear,

C C component starting at "first ,

316LSS Fatigue PEO (and earlier based on industry guidance) Primary inspection

Guide plates/cards CFS Weld, Wear,

C C component starting at first

Fatigue, TE PEO (and earlier based on CRGT industry guidance)

assemblies and flow Guide tube

SCC,IASCC, Component managed X-750 Wear, Fatigue, C C through plant-specific downcomers support pins

IE, ISR/IC programs

Upper Internals Assembly

No Additional Measures Sheaths 304 ss Wear, Fatigue C C component

--

,

UCP and fuel IASCC, Expansion inspection

Upper core plate 304 ss B C component starting at first alignment pins Fatigue, IE PEO

Upper Brackets, clamps,

instrumentation 304 ss No Additional Measures terminal blocks, SCC, Fatigue B C conduit and and conduit straps

302 ss component supports



July 9, 2018


Address on a plant-specific basis


- Evaluate need for contingency options · (See Table 6-5)



- Continue to implement plant-specific program


- Lead location is expected to be the welds to the CRGT flanges which are already an MRP-227 Primary inspection component. Therefore, the current inspection is adequate for asset management


- Consider performing proactive inspection prior to expansion criteria are triggered


- Consider performing expanded or supplemental inspections for asset management


I


Table 6-2: Westine:house Economic Risk Catee:orv C Component Dispositionine: (cont.) Screened-in

Safety Economic Current Asset Management Assembly Subassembly Component Material Degradation

Mechanisms Category Category Status

IASCC, Wear

Baffle-edge bolts 316 ss

(I), Fatigue, IE, B C Primary inspection component

347 ss starting at 20 to 40 EFPY

' ·, VS, ISR/IC

IASCC, Wear

Bracket Bolts 347 ss (I), Fatigue, IE, B C Primary inspection component

starting at 20 to 40 EFPY VS,ISR/IC

IASCC, Wear Primary inspection component

Baffle and Baffle-former bolts

316 ss (I), Fatigue, IE, C C

managed based on plant design

former 347 ss through MRP-227 interim

assembly VS, ISR/IC

guidance [101

IASCC, Wear Primary inspection component

- managed based on plant design Lower Comer Bolts 347 ss (1), Fatigue, IE, C C

through MRPs227 interim Internals - VS, ISR/IC

guidance no1 Assembly

IASCC, Wear Expansion inspection

Barrel-former bolts 316 ss

(I), Fatigue, IE, - B C component starting at first PEO

347 ss VS, ISR/IC

(and as managed through MRP-

227 interim guidance [11])

Lower core barrel Weld, IASCC, Expansion inspection

axial welds (MAW 304 ss B C

an\lLAW) Fatigue, IE, VS component starting at first PEO

Core Barrel

Lower core barrel

girth welds (LGW and 304 ss Weld, IASCC, B C


LFW) . Fatigue, IE, VS starting at first PEO


· Page 62 of70 LTR-AMLR-18-4, Rev. 1

July 9, 2018

Asset Management.Recommendation

- Current inspection adequate for asset

management - Evaluate need for contingency

options (See Table 6-5)


management

- Evaluate need for contingency


- Current inspection adequate for a~set

management



- Current inspection, adequate for asset

management



- Consider performing proactive

inspection prior to expansion criteria

are triggered



- Consider performing proactive

inspection prior to expansion criteria

are triggered*


options.(See Table 6-5)


management* - Evaluate need for contingency



Table 6-2: Westinghouse Economic Risk Categorv C Component Dispositioning (cont.) Screened-in

Safety Economic Current Asset Assembly Subassembly Component Material Degradation Mechanisms

Category Category Management Status

Upper core barrel Expansion inspection axial welds 304 ss Weld, Fatigue B C component starting at first

(UAW) PEO Core Barrel

Upper core barrel Primary inspection girth welds (UFW 304 ss Weld, Fatigue B C component starting at first

Lower andUGW) PEO Internals

Neutron Assembly panels/ Thermal shield 304 ss Primary inspection

thermal flexures 316 ss Weld, Fatigue B C component starting at first

shield PEO

Primary inspection Radial Radial support Stellite Wear C C component starting at first support keys keys

PEO

Primary inspection Clevis insert bolts X-750 SCC, Wear B C component starting at first

PEO

Primary inspection Clevis inserts Stellite Wear, Fatigue C C component starting 'at first

Interfacing Interfacing PEO

Components components Primary inspection Internals hold-

304 ss Wear, Fatigue B C component starting at first down spring PEO

Primary inspection

Thermal sleeves 304 ss Wear B C component managed based on plant design according to the industrv =idance f29l



July 9, 2018


- Consider performing proactive inspection prior to expansion criteria are triggered*





- Evaluate need for contingency optioris (See Table 6-5)



- Current inspection adequate for asset ,management








* Core barrel weld inspection requirements may be subject to change based on response to 2018 Spring outage operating experience and MRP-227, Rev. 1 safety evaluation


Table 6-3: Asset Management Inspection Recommendations for CE Components (NEI 03-08 Good Practice)

Asset Management Applicability Effect (Mechanism) Examination Method/ Frequency Item

Upper Internals All plants with welded Cracking (IASCC, Visual (VT-3) examination no later than

Assembly core shrouds assembled Fatigue), Aging 2 refueling outages from the beginning of

Fuel alignment plate in two vertical sections management (IE) the SLR period. Subsequent examination

on a ten-year interval.

Core Support Barrel All plants Cracking (SCC, IASCC, Enhanced visual (EVT-1) examination

Assembly Fatigue) Re-inspection every 10 years following

Upper cylinder axial Aging management (IE) initial inspection

welds(UAW)

Core Support Barrel All plants Cracking (SCC, IASCC, Enhanced visual (EVT-1) examination

Assembly Fatigue) Re-inspection every 10 years following

Lower cylinder axial Aging management (IE) initial inspection

welds (MAW and LAW)



July 9, 2018


Minimum of25% of core side surfaces

Minimum of75% of the OD length of the UAW and adjacent base metal

Minimum of75% of the OD length of the MAW and LAW and adjacent base metal



July 9, 2018

Table 6-4: Asset Management Inspection Recommendations for Westinghouse Components (NEI 03-08 Good Practice)

Asset Management . Applicability Effect (Mechanism) Examination Method/ Frequency Examination Coverage Item·

Upper Internals All plants Cracking (IASCC, Visual (VT-3) examination no later than Minimum of25% of core side Assembly Fatigue), Aging 2 refueling outages from the beginning of surfaces

Upper core plate management (IE) the SLR period. Subsequent examination on a ten-year interval.

\

Upper Instrumentation All plants Cracking (SCC, Fatigue) Visual (VT-3) examination no later than 100% of accessible brackets, clamps, Conduit and Supports 2 refueling outages from the beginning of terminal blocks, and conduit straps in

Brackets, clamps, the SLR period. Subsequent examination peripheral CRGT assemblies (i.e.,

terminal blocks, and on a ten-year interval. those adjacent to the perimeter of the

conduit straps upper core plate)

Core Barrel Assembly All plants · Cracking (IASCC, Visual (VT-3) examination no later than 100% of accessible bolts.

Barrel-former bolts Fatigue) 2 refueling outages· from the· beginning of

Aging Management (IE the SLR period. Subsequent examination

'

and ISR) on a ten-year interval.

Core Barrel Assembly All plants Cracking (SCC, Fatigue) Enhanced visual (EVT-1) examination Minimum of75% of the OD length of

Upper core barrel axial Re-inspection every 10 years following the UAW and adjacent base metal

welds (UAW) initial inspection

Core Barrel Assembly All plants Cracking (SCC, IASCC, Enhanced visual (EVT-1) examination Minimum of75% of the OD length of

Lowercore barrel axial Fatigue) Re-inspection every 10 years following

the MAW and LAW and adjacent base

welds (MAW and LAW) Aging Management (IE) initial inspection metal

'c Distortion (VS)

']


Page 66 of70 LTR-AMLR-f8-4, Rev. I

July 9, 2018

Table 6-5: High-Level Options and Supplemental Resources for the Development oflnspection Contm2encv Ootions for Asset Mana ?:ement ·

Monitor: - Follow operating experience - Loose parts monitoring - Neutron noise monitoring - Measurement channel integrity monitoring - Other

Mitigate: - Change fuel loading or core design - Remove broken parts - Peening - Other

High-Level Contingency Options Replace:

Supplemental Resources (Note that this is not guaranteed to be an all-encompassing list of resources useful for developing contingency options)

Repair:

Full upper or lower internals replacement Acceptable pattern replacement Sub-assembly replacement Other

- Welding - Supplemental support device - Crack arresting - Other

No Additional Measures

Inspection Response Plans - Generic risk ra_pking and response planning results

(WCAP-17436-P [ 42]) - Plant-specific reports

MRP-318 [43]

- Expert panel review of strategy, prioritization, and contingency options for PWR vessel internals

WCAP-17096-NP-A [44]

- Reactor internals acceptance criteria methodology and data requirements

WCAP-17451-P [36]

- CRGT guide card wear guidance

PWROG-17055-NP [ 45] - Contingencies for mitigating guide card wear

MRP 2017-013 [46]

- Baffle-Former Bolt Playbook

MRP 2017-034 [47] - Clevis Insert Bolt Playbook

7 Conclusions I



July 9, 2018

Development of the MRP-227 revision for SLR is a process with multiple steps that is not expected to be finished until the end of 2020. However, the utility owners of the lead plants for SLR are already preparing SLR applications and require the best possible information available on the expected I&E guidelines for the reactor vessel internals. This interim guidance letter references the best information currently available about how the future I&E guidelines may change when operation into the second PEO is considered.

Any application of these results must recognize that this information is based on the information available as of the date of publication. Some of that information is preliminary, such as the results of the SLR expert panel review. Some of that information may be modified by future analyses or evaluations or even operating experience that occurs between publication of this letter and publication of the MRP-227 revision for SLR. Note that there may be some conservatism in the results presented here because all of the high risk items from the SLR expert panel review were included and addressed in some way. Several were added as MRP-227 Primary, Expansion, or Existing inspection components.

Application of these results µrnst also confirm applicability of the criteria provided in Section 2.

The tables and supporting text in Section 5 can be used in developing an SLR aging management program for the reactor vessel internals. These sections use MRP-227-A as a starting point and provide direction on what guidance from MRP-227, Revision 1 should be included in an aging management program. It is recommended that applications basing a reactor vessel internals aging management program on this interim guidance also inclµde a commitment to actively monitor industry development of the final l&E guidelines for SLR and to reconcile differences between the interim gui~ance provided here and the final I&E guidelines.

The SLR expert panel provided direction on both the safety-related risks for reactor vessel internals components and the economic risks. Section 6 provides direction for several components that received a high economic risk category during the expert panel review but are not addressed through MRP-227 Primary or Existing inspection requirements. This asset management guidance can be used to help support long-term commercial viability of a plant.

'\


8 References



July 9, 2018

l. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

2. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision I). EPRI, Palo Alta, CA: 2015. 3002005349.

3. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision I). EPRI, Palo Alto, CA: 2017. 3002010268,

4. Materials Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanism, Models, and Basis Data-State of Knowledge (MRP-211, Revision I). EPRI, Palo Alto, CA: 2017. 3002010270.

5. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191, Revision I). EPRI, Palo Alto, CA: 2016. 3002007960.

6. Westinghouse Letter, LTR-AMLR-17-28, Revision 1, "MRP-191 Revision 2 Expert Panel Meeting Minutes," J~ne ·15, 2018 (Westinghouse Proprietary).

7. U.S. Nuclear Regulatory Commission Report, NUREG-2191, "Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report," July 2017 (ADAMS Accession Number MLI 7f87A031).

8. U.S. Nuclear Regulatory Commission Report, NUREG-2192, "Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants," July 2017 (ADAMS Accession Number MLl 7188A 158).

9. Guidelines for the Management of Materials Issues, NEI 03-08, Revision 3, Nuclear Energy Institute, Washington, DC, February 2017.

10. MRP Letter MRP 2017-009, "Transmittal ofNEI-03-08 "Needed" Interim Guidance Regarding Baffle Former Bolt Inspections for PWR Plants as Defined in Westinghouse NSAL 16-01 Rev. 1," March 15, 2017. (ML17087A106)

11. MRP Letter MRP 2018-002, "Transmittal ofNEI 03-08 "Needed" Interim Guidance Regarding MRP-227-A and MRP-227; Revision 1 Baffle-Former Bolt Expansion Inspection Requirements for PWR Plants," January 17, 2018.

12. Pressurized Water Reactor Owners Group Letter, OG-18-46, "Transmittal of Approved "Needed" Interim Guidance for Addressing Accelerated Guide Card Wear Issue Described in NSAL-17-1, (LTR-ARIDA-17-270, Revision 0), PA-MSC-1471," February 20, 2018.

13. Pressurized Water R~actor Owners Group Letter, OG-18-76, "Transmittal oflnterim Guidance for Addressing Accelerated Guide Card Wear Issue Described in NSAL-17-1 (PA-MSC-14 71 ),"

. March 23, 2018. (ML18088A199) 14. MRP Letter MRP 2017-027, Revision 0, "Responses to NRC Request for Additional Information

for Electric Power Research Institute Topical Report MRP-227, Revision 1, 'Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline' (CAC No. MF7740)," October 16, 2017. (ML17305A056) ·

15. MRP Letter, MRP 2018-003, "Responses to NRC Request for Additional Information for Electric Power Research Institute Topical Report MRP-227, Revision 1, "Materials Reliability Program:



July 9, 2018

Pressurized Water Reactor Internals Inspection and Evaluations Guideline" (CAC No. MF7740)," January 30, 2018. (ML18038A875)

16. MRP Letter MRP 2018-011, Revision 0, "Transmit Supplemental Information Regarding EPRI --' Technical Report MRP-227-Revision 1, 'Materials Reliability Program: Pressurized Water

Reactor Internals Inspection and Evaluation Guidelines,'" May 17,2018. (ML18142A233, ML18142A234, ML18142A234, ML18142A235, ML18142A236, and ML18142A237)

17. Materials Reliability Program Letter, MRP 2013-025, "MRP-227-A Applicability Template Guideline," October 14, 2013.

18. Westinghouse Calculation, CN-REA-17-23, Revision 0, "EPRI MRP-191 Fluence Evaluation for Combustion Engineering Reactor Internals," Aug1,1st 8, 2017. (Westinghouse Proprietary)

19. Westinghouse Calculation, CN-REA-17-46, R~vision 2, "EPRI MRP-191 Fluence Evaluation for Westinghouse Reactor Internals," December 14, 2017. (Westinghouse Proprietary)

20. U.S. Nuclear Regulatory Commission Safety Evaluation, "Final Safety Evaluation of Action Items 1 and 7 from Topical Report MRP-227-A, 'Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline,' (CAC No. MF7223 EPID: L

-2016-TOP-0001)," January 29, 2018. (ML18016A008) 2 l. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism

Screening and Threshold Values (MRP-175, Revision OJ. EPRI, Palo Alto, CA: 2005. 1012081. 22. Westinghouse Letter, LTR-AMLR-17-22, Revision 0, "MRP~191 for Subsequent License

Renewal Expert Panel Screening Inputs," January 18, 2018. (Westinghouse Proprietary) 23. 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.

24. Westinghouse Nuclear Safety Advisory Letter, NSAL-10-1, "Rod Control Cluster Assembly Guide Card Wear," March 9, 2010.

25. Materials Reliability Program: Inspection Data Survey Report (MRP-219, Revision 11). EPRI, Palo Alto, CA: 2015. 3002005509.

26. Westinghouse Nuclear Safety Advisory Letter, NSAL-17-1, "Guide Tube Guide Card Wear I

Attributed to Ion Nitride Rod Cluster Control Assembly," January 16, 2017. 27. Westinghouse Technical Bulletin, TB-12-5, "Baffle Bolt Degradation in a Westinghouse NSSS

Plant with Downflow Reactor Internal Design," March 7, 2012. 28. Westinghouse Nuclear Safety Advisory Letter, NSAL-16-1, Revision 1, "Baffle-Former Bolts,"

August 1, 2016. 29. Westinghouse Technical Bulletin, TB-07-2, Revision 3, "Reactor Vessel Head Adapter Thermal

Sleeve Wear," December 7, 2015. 30. MRP Letter MRP 2018-010, "Notification of Recent PWR CRDM Thermal Sleeve Flange Wear

and Control Rod Motion Stoppage Operating Experience and Recommended Plant Actions,". April 20, 2018.

31. World Associatiori of Nuclear Operators Event Report, 2018-0110, "Blockage of a Control Rod Due to Wear in the Thermal Sleeves of Vessel Heads in the 1300MW and N4 Series Fleet

J

Belleville 2, 13/12/2017 Noteworthy," March 1, 2018. 32. Nuclear Regulatory Commission Event Report, Event Number 53422, "Part 21 - Wear of

Thermal Sleeve in Westinghouse Control Rod Drive Mechanisms," May 23, 2018.

Westinghouse Non'...Proprietary Class 3 Enclosµre to MR.p 2018-02~


July 9, 2018

33. Westinghouse Technical Bulletin, TB-14-5, "Reactor Internals Lower Radial Support Clevis

Insert Cap Screw Degradation," August 25, 2014.

34. Westinghouse Technical Bulletin, TB-16-4, "Fuel Alignment Pin Malcomized Surface

Degradation," August 15, 2016 .. 35. Garner, F.A., "New Insights on the Potential Void Swelling of AISI 304 and 316 Stainless Steels

in the PWR Baffle-Former Assembly," International Light Water Reactor Materials Reliability

Conference and Exhibition, Chicago, Illinois, August 1-4, 2016.

36. Westinghouse Report, WCAP-17451-P, Revision 1, "Reactorlnternals Guide Tube Wear

Westinghouse Domestic Fleet Operational Projections," October 2013 (Westinghouse

Proprietary). 37. U.S. Nuclear Regulatory Commission Document, "Office of Nuclear Reactor Regulation Staff

Assessment of Electric Power Research Institute NE 03-08, Revision 2 "Needed" Interim

Guidance Regarding Baffle-Former Bolt Inspections in Westinghouse-Design Pressurized Water

Reactors," November 20, 2017 (ADAMS Accession No. ML 1731 OA86 l ).

38. Pressurized Water Reactor Owners Group Report, PWROG-15034-P, Revision 0, "Clevis Bolt

Fa~rication and Inspection Assessment," January 2016 (Proprietary).

39. MRP Letter, MRP-2014-006, "Materials Reliability Program: Pressurized Water Reactor

Internals Inspection and Evaluation Guidelines (MRP-227-A), EPRI, Palo Alto, CA: 2011.

1022863, Transmittal oflnterim Guidance," February 18, 2014. (ML14274A372)

40. Westinghouse Report WCAP-16911-P, Revision 0, "Reactor Vessel Head Thermal Sleeve Wear

Evaluation for Westinghouse Domestic Plants," July 2008 (Westinghouse Proprietary).

41. Pressurized Water Reactor Owners Group Report, PWROG-16003-P, Revision 1, "Evaluation of

Potential Thermal Sleeve Flange Wear," August 2017 (Proprietary).

42. Pres~urized Water Reactor Owners Group Report, WCAP-17436-P, Revision 0, "Reactor

Internals Risk Ranking and Response Planning Results for PWROG Project PA-MSC-0568,"

November 2011 (Proprietary).

43. Materials Reliability Program: Proceedings of the Expert Panel on Strategy Review,

Prioritization and Contingency Planning Options for PWR Internals Components (MRP-318).

EPRI, Palo Alto,CA: 2011. 10228()4.

44. Pressurized Water Reactor Owners Group Report, WCAP-17096-NP-A, Revision 2, "Reactor

Internals Acceptance Criteria Methodology and Data Requirements," August 2016.

45. Pressurized Water Reactor Owners Group Report, PWROG-17055-NP, Revision 0, "List of

Applicable Contingencies for Mitigating Guide Card Wear," November 2017.

46. Materials Reliability Program Letter, MRP 2017-013, Revision 0, "Transmit Baffle-Former-Bolt

Response Options 'Playbook' for Information and Use in Planning for PWR Refueling Outages,"

May 12, 2017. (ML17222A167)

47. Materials Reliability Program Letter, MRP 2017-034, Revision 0, "Generic PWR Clevis Insert

Cap Screw (Bolting) Response Option Playbook," December 22, 2017.

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