PWROG-17071-NP, 'WCAP-17096-NP-A Interim Guidance.'participating utility representatives of the Materials Committee. This document is a deliverable under PWROG Project Authorization

• •~~~~~~-=--~-~~--~~--""""""""""""""'~~~-"""""""'-• • • • • • • • • • •

PWROG-1707,t .. 1,f P ... Revision 0

• ~~~~~~~iifiliilillilliilll!llill,ll/;!limffli!~-l!!Hliill::JIIGliiD~~lllI!lilUI!~~~--.,~~~~~

• • • • • • • • • • • • :, •• • • • • • • • • • • • • a

WESTINGHOUSE NON-PROPRIETARY CLASS 3

WCAP-17096-NP-A Interim Guidance Materuais Comm~ttee

PA~MSC~1567

March 2018

. . ---~ - " ~- ----ii....-- ::::::L-=----------~-- .. ~ . - . =• - '"' '• ~ - ...

- -· ~ ~

••• This record was final approved on 3/13/2018 2:13:14 PM. ( This statement was added by the PRIME system upon its validation)


PWROG-17071-NP Revision 0

WCAP-17096-NP-A Interim Guidance

PA-MSC-1567

David C. Radonovich* and Bradley T. Carpenter* Aging Management & License Renewal

Sarah Davidsaver** and Steve Fyfitch**, Sections 1.2 and 2.1 Framatome

March 2018

Reviewer: Matthew J. Palamara* Reviewer: John F. Kielb* Aging Management & License Renewal Reactor Internals Design & Analysis - II

Approved: John L. McFadden, Manager* Approved: James P. Molkenthin, Program Director* PWR Owners Group PMO Aging Management & License Renewal

This document may contain technical data subject to the export control laws of the United States. In the event that this document does contain such information, the Recipient's acceptance of this document constitutes agreement that this information in document form (or any other medium), including any attachments and exhibits hereto, shall not be exported, released or disclosed to foreign persons whether in the United States or abroad by recipient except in compliance with all U.S. export control regulations. Recipient shall include this notice with any reproduced or excerpted portion of this document or any · document derived from, based on, incorporating, using or relying on the information contained in this

· document.

This document is the property of and contains Proprietary Information owned by Westinghouse Electric Company LLC and/or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document in strict accordance with the terms and conditions of the agreement under which it was provided to you. Any unauthorized use of this document is prohibited.

*Electronically approved records are authenticated in the electronic document management system.

**The hand signatures of the Framatome authors are contained in the referenced Framatome document.

PWROG-17071 (3-6-18).docx-030618

Westinghouse Electric Company LLC 1000 Westinghouse Drive

Cranberry Township, PA 16066, USA

Framatome Inc. 3315 Old Forest Road Lynchburg, VA 24501

© 2018 Westinghouse Electric Company LLC All Rights Reserved

*** This record was final approved on 3/13/2018 2:13:14 PM. ( This statement was added by the PRIME system upon its ,validation)

• • • • • • • • • • • • • • • • • • • • • • • • i

• • • ., •! .I • • • • • • .I • • •1 ., •I ~

•• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •


ACKNOWLEDGEMENTS

This report was developed and funded by the PWR Owners Group under the leadership of the participating utility representatives of the Materials Committee. This document is a deliverable under PWROG Project Authorization PA-MSC-1567 .

iii

PWROG-17071-NP March 2018 Revision 0

*** This record was final approved on 3/13/2018 2:13:.14 PM. (,This statement was added by the PRIME system llpon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv

LEGAL NOTICE

This report was prepared as an account of work performed by Framatome Inc. and Westinghouse Electric Company LLC. Neither Framatome, nor Westinghouse Electric Company LLC, nor any person acting on its behalf:

1. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

2. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.

COPYRIGHT NOTICE

This report has been prepared by Framatome Inc. and bears an Framatome Inc. copyright notice. Information in this report is the property of, and contains copyright material owned by, Framatome Inc. and /or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document and the material contained therein in strict accordance with the terms and conditions of the agreement under which it was provided to you.

This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company copyright notice. Information in this report is the property of and contains copyright material owned by Westinghouse Electric Company LLC and /or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to

. treat this document and the material contained therein in strict accordance with the terms and conditions of the agreement under which it was provided to you.

As a participating member of this task, you are permitted to make the number of copies of the information contained in this report that are necessary for your internal use in connection with your implementation of the report results for your plant(s) in your normal conduct of business. Should implementation of this report involve a third party, you are permitted to make the number of copies of the information contained in this report that are necessary for the third party's use in supporting your implementation at your plant(s) in your normal conduct of business if you have received the prior, written consent of Westinghouse Electric Company LLC to transmit this information to a third party or parties. All copies made by you must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.



• • • • • • • • • • • • • • • • • • • • • • • •1 • • • • • • • • • • • • I • • • .I ~ • ·•-

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V

DISTRIBUTION NOTICE

This report was prepared for the PWR Owners Group. This Distribution Notice is intended to establish guidance for access to this information. This report (including proprietary and non-proprietary versions) is not to be provided to any individual or organization outside of the PWR Owners Group program participants without prior written approval of the PWR Owners Group Program Management Office. However, prior written approval is not required for program participants to provide copies of Class 3 Non-Proprietary reports to third parties that are supporting implementation at their plant, and for submittals to the NRG .


*** This record was final approved on 3/13/2018 2:13:14 PM. ( This statement was added by the P,RIME system upon its validation)


PWR Owners Group United States Member Participation* for PA-MSC-1567

Utility Member Plant Site(s)

Ameren Missouri Callaway (W)

American Electric Power D.C. Cook 1 & 2 (W)

Arizona Public Service Palo Verde Unit 1, 2, & 3 (CE)

Dominion Connecticut Millstone 2 (CE)

Millstone 3 (W)

North Anna 1 & 2 (W) Dominion VA

Surry 1 & 2 (W)

Catawba 1 & 2 (W)

Duke Energy Carolinas McGuire 1 & 2 (W)

Oconee 1, 2, & 3 (B&W)

Robinson 2 (W) Duke Energy Progress

Shearon Harris (W)

Entergy Palisades Palisades (CE)

Entergy Nuclear Northeast Indian Point 2 & 3 (W)

Arkansas 1 (B&W)

Entergy Operations South Arkansas 2 (CE)

Waterford 3 (CE)

Braidwood 1 & 2 (W)

. Byron 1 & 2 (W) ·

Exelon Generation Co. LLC TMI 1 (B&W)

Calvert Cliffs 1 & 2 (CE)

Ginna (W)

FirstEnergy Nuclear Operating Co. Beaver Valley 1 & 2 (W)

Davis-Besse (B&W)

St. Lucie 1 & 2 (CE)

Florida Power & Light\ NextEra Turkey Point 3 & 4 (W)

Seabrook (W)

Pt. Beach 1 & 2 (W)

PWROG-17071-NP

vi

Participant

Yes No

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

March 2018 Revision 0

*** This record was final approved on 3/13/2018 .2:13:14 PM. ( This statement was added by the PRIME system upon its validation)

• • • • • I

•i ., • • ., el

I ., • ., el

I • • • • I

• ., •1 • • • • • • .i • • • • • • • • • • •1 • • ••

• • • • • • • • • • • • • • • • • • • • • • • • ~-• • • • • • • • • • • • • • • • • •


PWR Owners Group United States Member Participation* for PA-MSC-1567

Participant

Utility Member Plant Site(s) Yes No

Luminant Power Comanche Peak 1 & 2 (W) X

Omaha Public Power District Fort Calhoun (CE) X

Pacific Gas & Electric Diablo Canyon 1 & 2 (W) X

PSEG - Nuclear Salem 1 & 2 (W) X

South Carolina Electric & Gas V.C. Summer (W) X

So. Texas Project Nuclear Operating Co. South Texas Project 1 & 2 (W) X

Farley 1 & 2 (W) X Southern Nuclear Operating Co.

X Vogtle 1 & 2 (W)

Sequoyah 1 & 2 (W) X Tennessee Valley Authority

Watts Bar 1 & 2 (W) X

Wolf Creek Nuclear Operating Co. Wolf Creek (W) X

Xcel Energy Prairie Island 1 & 2 (W) X

* Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above .

vii


*** This record was final approved on 3/13/2018 2:13:14 PM. (This.statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii

PWR Owners Group

nternat1ona I M b P f * f PA MSC 1567 em er art1c1pa 10n or - -Participant

Utility Member Plant Site(s) Yes No

Asco 1 & 2 (W) X Asociaci6n Nuclear Asc6-Vandell6s

Vandellos 2 (W) X

AxpoAG Beznau 1 & 2 (W) X

Centrales Nucleares Almaraz-Trillo Almaraz 1 & 2 (W) X

EDF Energy Sizewell B (W) X

Electrabel Dael 1, 2 & 4 (W) X

Tihange 1 & 3 (W) X

Electricite de France 58 Units X

Elektriciteits Produktiemaatschappij Zuid-Nederland Borssele 1 (Siemens) X

Eletronuclear-Eletrobras Angra 1 (W) X

Emirates Nuclear Energy Corporation Barakah 1 & 2 X

Eskom Koeberg 1 & 2 (W) X

Hokkaido Tomari 1, 2 & 3 (MHI) X

Japan Atomic Power Company Tsuruga 2 (MHI) X

Mihama 3 (W) X

Kansai Electric Co., LTD Ohi 1, 2, 3 & 4 (W & MHI) X

Takahama 1, 2, 3 & 4 (W & MHI) X

Kori t, 2, 3 & 4 (W) X

Hanbit 1 & 2 (W) X Korea Hydro & Nuclear Power Corp.

X Hanbit 3, 4, 5 & 6 (CE)

Hanul 3, 4 , 5 & 6 (CE) X

Genkai 2, 3 & 4 (MHI) X Kyushu

Sendai 1 & 2 (MHI) X

Nuklearna Electrarna KRSKO Krsko (W) X

RinghalsAB Ringhals 2, 3 & 4 (W) X

Shikoku lkata 1, 2 & 3 (MHI) X

Taiwan Power Co. Maanshan 1 & 2 (W) X

* Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.


*** This reco_rd was final approved on 3/13/2018 2:13:14 PM. ( This statement was added by the PRIME system upon its validation)

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •

• • • • • • • • • • • • • ,. • • :e • ,. • • • • • • • • • • • • • • • • • • • • • • • •

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix

TABLE OF CONTENTS

1 SCOPE OF CHANGES ..................................................................................................... 1 1.1 BACKGROUND/ PURPOSE ................................................................................ 1 1.2 B&W PLANT DESIGN [11] .................................................................................... 2 1.3 COMBUSTION ENGINEERING PLANT DESIGN ................................................. 2

1.3.1 MRP-227 Component Cross Reference Matrix ...................................... 2 1.3.2 Weld Crack Growth Rate ........................................................................ 3

1.4 WESTINGHOUSE PLANT DESIGN ...................................................................... 3 1.4.1 MRP-227 Component Cross Reference Matrix ...................................... 3 1.4.2 · Baffle-Former Bolt Guidance .................................................................. 3 1.4.3 Guide Card Wear Guidance ................................................................... 4 1.4.4 Miscellaneous Changes .......................................................................... 4

1.5 CONCLUSION ...................................................................................................... 5 2 INTERIM GUIDANCE ....................................................................................................... 6

2.1 B&W PLANT DESIGN [11] .................................................................................... 6 2.1.1 Original and Modified Vent Valve Locking Devices ................................. 6 2.1.2 Vent Valve Bodies ................................................................................... 9 2.1.3 Lower Grid Rib Section ......................................................................... 11

2.2 COMBUSTION ENGINEERING PLANT DESIGN ............................................... 13 2.2.1 Cross Reference Matrix ........................................................................ 13 2.2.2 Weld Crack Growth Rate ...................................................................... 15

2.3 WESTINGHOUSE PLANT DESIGN .................................................................... 16 2.3.1 Cross Reference Matrix ........................................................................ 16 2.3.2 Baffle-former Bolts ................................................................................ 18 2.3.3 Thermal Shield Flexure ......................................................................... 21 2.3.4 Weld Crack Growth Rate ...................................................................... 22

3 REFERENCES .... : ........... , .................................. , .................... : ..................................... ,.23


*** This record was final approved on 3/13/2018 2:13:14 PM. ( This statement was added by the PRIME system upon its validation)

• • • • • • • • • • • • • • • • • • • • • • ,. • • • • • • • • • • • • • • • • • • • •


1 SCOPE OF CHANGES

1.1 BACKGROUND / PURPOSE

The purpose of MRP-227 is to provide inspection and evaluation guidelines for reactor vessel internals (RVI) in pressurized water reactors. MRP-227 inspections are required by aging management plans (AMPs) as part of first license renewal for domestic plants. The methodology and data requirements for developing acceptance criteria for RVI inspection are provided in WCAP-17096-NP-A, Rev. 2 [2]. Both of these documents required review and approval by the Nuclear Regulatory Commission (NRC) to gain acceptance. In addition, the initial MRP-227. inspections have occurred at many plants, and the lessons learned from the operating experience must be incorporated .

1

The original version of MRP-227 (i.e. MRP-227-A) was reviewed and approved by the NRC, which has since been updated and submitted to the NRC as MRP-227, Revision 1 [1]. The NRC provided several Requests for Additional Information (RAls) [3] and the industry responses to these are currently in various stages of development and issuance. WCAP-17096-NP-A, Rev. 2 [2] received NRC approval, but other industry initiatives and operating experience (OE) have occurred since and necessitate a revision to the document.

The purpose of this document is to provide interim guidance to the industry related to anticipated changes to the acceptance criteria methodology and data requirements provided in in WCAP-17096-NP-A, Rev. 2 [2]. The expected updates to the document are due to the release of industry guidance documents and updates of existing documentation, i.e. MRP-227, Rev. 1 [1], and to incorporate lessons learned from the MRP-227 inspections performed to date .

In order to generate this interim guidance, WCAP-17096-NP-A, Rev. 2 was reviewed to identify areas where the current guidance is inconsistent with other industry issued guidance documents that have been issued since the NRC approval of WCAP-17096-NP-A. The changes must be substantive to be included in the interim guidance, so this does not include clarification and consistency type changes which will be incorporated into the next revision of WCAP-17096 .

. With consideration of this approach, interim guidance on WCAP-17096-NP-A, Rev. 2 will be provided for the following items as described herein .



WESTINGHOUSE NON-PROPRIETARY CLASS 3 2

1.2 B&W PLANT DESIGN [11]

This interim guidance does not contain any technical1 updates to information (e.g., methodologies) for component items and welds already in WCAP-17096-NP-A per MRP-227-A. The only items in this interim guidance are component items and welds that are not in MRP-227-A, but are in MRP-227, Rev. 1, both Primary and Expansion. These component items and welds are:

1. Original and modified vent valve locking devices

2. Vent valve bodies

3. Lower grid rib section

1.3 COMBUSTION ENGINEERING PLANT DESIGN

1.3.1 MRP-227 Component Cross Reference Matrix

A cross reference matrix is provided in Section 2.2.1 to clarify the requirements for each respective component, which includes the changes that were made in updating MRP-227-A to MRP-227, Rev. 1 [1], identification of the appropriate methodology, and any applicable RAls from the NRC review of Revision 1 of MRP-227.

This item is necessary due to the significant number of changes that have occurred in updating MRP-227-A to MRP-227, Rev. 1 [1], which include component identification numbering, component nomenclature, category (i.e., primary or expansion), realignment of primary and expansion links, plant applicability, degradation effects and failure mechanisms, applicable examination methods, and examination coverage. In addition, certain components, such as the upper core barrel flange (CE-ID: 6.3 as defined in [2]) have been eliminated from MRP-227 inspections. Despite these changes, the methodologies defined in WCAP-17096-NP-A, Rev. 2 [2] general.ly remain applicable with exception ofthe weld crack growth rate as discussed below. In addition, the.resolutionof RAls [3] must be monitored for potential changes;

1 The tables in Appendix A of WCAP-17096-NP-A, Rev. 2 [2], which are repeated from MRP-227-A for each component item and weld, will be removed to avoid version issues and potential copy errors between.the two documents. PWROG-17071-NP March 2018

Revision 0


• • • • • • • • • • • • • • • • • • • el • • ., •1 .: • • • • • • !

• • • . : • • • • • • • •

• • • • • • • • i. i. • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •


1.3.2 Weld Crack Growth Rate

A specific EPRI reference is available for the crack growth rate in [9] for welds with fluence greater than 3x1021 n/cm2

. This revised guidance applies to the methodology defined for the following components in WCAP-17096-NP-A, Rev. 2[2]:

o CE-ID: 2 - Core Shroud Assembly (Welded) - Core Shroud Plate-Former Plate Weld o CE-ID: 2.1 - Core Shroud Assembly (Welded) - Remaining Axial Welds o CE-ID: 3 - Core Shroud Assembly (Welded - Full Height) - Shroud Plates

3

o CE-ID: 3.1 - Core Shroud Assembly (Welded)- Remaining Axial Welds, Ribs and Rings o CE-ID: 7 - Core Support Barrel Assembly - Lower Cylinder Girth Welds o CE-ID: 7.1 - Core Support Barrel Assembly - Core Barrel Assembly Axial Welds

Note that a Section XI ASME Code Case is in development on this topic, which is expected to become applicable upon completion .

1.4 WESTINGHOUSE PLANT DESIGN

1.4.1 MRP-227 Component Cross Reference Matrix

A cross reference matrix is provided in Section 2.3.1 to clarify the requirements for each respective component, which includes the changes that were made in updating MRP-227-A to MRP-227, Rev. 1 [1], identification of the appropriate methodology, and the applicable RAls from the NRC review of MRP-227, Rev. 1 .

This task is necessary due to the significant number of changes that have occurred in updating MRP-227-A to MRP-227, Rev. 1 [1], which include component identification numbering, component nomenclature, category (i.e., primary or expansion), realignment of primary and expansion links, plant applicability, degradation effect and failure mechanism, applicable. examination method, and examination coverage. -In addition, certain components, such as the

· core barrel outlet nozzle welds (W'."'ID: 3.1 as defined in [2]) have been eliminated from MRP-227 inspections. Despite these changes, the.m~thodologies defined iri WCAP-17096"'.NP-A, Rev. 2 [2] generally remain applicable with exception of the baffle-former bolts, and the rniscellaneous items described below. In addition, the resolution of RAls [3] must be monitored for potential changes .

1.4.2 Baffle-Former Bolt Guidance

Interim guidance has been provided for baffle-former bolts in Section 2.3.2, which will supersede that currently defined by the methodology in WCAP-17096-NP"'."A, Rev. 2. This is necessary due to lessons learned from recent baffle-former bolt inspections, which resulted in the interim guidance provided in MRP-2016-021 [5] and MRP 2017-009 [4]. These documents provide the initial inspection and re-inspection intervals for Westinghouse plants based upon the recent operating experience (OE) with baffle-former bolt degradation, which differs from that originally defined in WCAP-17096-NP-A, Rev. 2 [2] .




The previous methodology for the baffle-former bolt analysis and acceptance criteria was defined in W-10: 7 of WCAP-17096-NP-A, Rev. 2 [2] (or W6 of MRP-227, Rev. 1 [1]). This interim guidance applies the criteria from MRP-2017-009 [4] (i.e., percent indications and clustering) to categorize the UT inspection findings. A plant-specific evaluation is required for plants with accelerated degradation, where the acceptable pattern must include margin to account for the degradation rate. Note that a valid model of degradation rate must account for failure initiation due to the IASCC mechanism, which is a function of plant-specific design parameters such as stress, temperature, and fluence.

This evaluation will account for the as-found condition and/or bolt replacements as applicable, plant-specific design and operating parameters, and apply a method for demonstrating acceptable bolting patterns such as that defined in WCAP-15029-P-A [6]. Plants with UT inspection results below the indication and clustering criteria threshold in MRP-2017~009 [4] will have the option to continue to apply the 50% margin consumed criteria defined in WCAP-17096-NP-A, Rev. 2 [2].

1.4.3 Guide Card Wear Guidance

Guide card wear outside of the anticipated limits specified in WCAP-17451-P, Rev. 1 [10] was noted on plants with ion nitride rod cluster control assembly (RCCA) guide tubes. The affected guide tubes includes those with 17x17 A or 17x17 AS style guide tubes, which include four plants at U.S. utilities [8]. This issue has been submitted to the NRC as a defect pursuant to 1 O CFR Part 21.

Interim guidance on the MRP-227 inspection of guide cards, as defined in WCAP-17451-P, was issued per OG-18-46 [15]. These changes impact the baseline inspection intervals and sample size guidance defined for the guide cards (W-10: 1) in WCAP-17096-NP-A, Rev. 2 [2]. Therefore, the baseline inspection interval and sample size guidance defined in WCAP-17096-NP-Awill be updated to include the necessary changes. The WCAP-17096-NP-A, Rev. 2 Interim Guidance for this component will be limited to clarification on baseline inspections, as per the WCAP-17451-P,-Rev. 1 Interim Guidance.

1.4.4 Miscellaneous Changes

A few other items require update, which are considered miscellaneous changes as discussed in the following.

Thermal Shield Flexure Fatigue Assessment

The methodology used for the fatigue evaluation of the thermal shield flexures (W-1 D: 10 in WCAP-17096 [2] or W9 in MRP-227, Rev. 1 [1]) has been updated to remove unnecessary constraints in the development of the acceptance criteria. As shown in Section 2.3.3, Steps 5 and 6 will be updated to require a "fatigue assessment" rather than explicit calculation of "fatigue usage" of less than 1.0. Likewise, "Increased loading on the core barrel wall should also be addressed" will be removed. This change allows flexibility for interpreting the necessary steps



• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •

• • • • • • • • • • • ., • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •


without requiring undue supplemental scope. This is an acceptable change since a limited quantity of transient cycles could occur until the next inspection .

Weld Crack Growth Rate

A specific EPRI reference is available for the crack growth rate in [9] for welds with fluence greater than 3x1021 n/cm2

• This revised guidance applies to the methodology defined for the following components in WCAP-17096, Rev. 2-A [2]:

o W-ID: 4 - Core Barrel Assembly - Upper and Lower Core Barrel Cylinder Girth Welds o W-ID: 4.1 - Core Barrel Assembly - Upper and Lower Core Barrel Cylinder Axial Welds

Note that a Section XI ASME Code Case is in development on this topic, which is expected to become applicable upon completion .

1.5 CONCLUSION

WCAP-17096-NP-A, Rev. 2 has been reviewed to determine the changes necessary to B&W, Combustion Engineering, and Westinghouse designed plants for inclusion in Interim Guidance to be issued for industry use while WCAP-17096-NP-A, Rev. 2 is being revised and evaluated for potential submittal to the NRC for review and approval. These requirements have been identified as part of Task 2 of PA-MSC-1567, and are modified in this report to provide interim guidance to participating utilities .

5




2 INTERIM GUIDANCE

2.1 B&W PLANT DESIGN [11]

2.1.1 Original and Modified Vent Valve Locking Devices

Problems were noted involving the original locking devices for the B&W-design vent valve jackscrews in the late 1970s and early 1980s. The jackscrew locking mechanism was vibrating and wearing through the locking cup. A new locking mechanism was designed and supplied to most B&W units. Additionally, several crimped locking cups in the vent valves have been found with crack-like indications at several units.

Component Item Function:

6

Vent valve assemblies are installed in the mounting rings in the core support shield. For all normal operating conditions, the vent valve is closed. In the event of a rupture of the reactor vessel inlet pipe, the valve will open to vent steam generated in the core directly to the break, thus permitting the core to .be flooded and adequately cooled after emergency core coolant has been supplied to the reactor vessel. Each vent valve assembly consists of a hinged disc, a valve body with sealing surfaces, a split retaining ring and fasteners, and an alignment device (to maintain the correct orientation). Each vent valve assembly can be remotely handled as a unit for removal or installation. Vent valve component parts, including the disc, are designed to minimize the possibility of loss of parts to the reactor coolant system, and all operating fasteners include a positive locking device. The hinged-disc includes a device for remote testing and verification of proper disc function. The external side of the disc is contoured to absorb the impact load of the disc on the RV inside wall without transmitting excessive impact loads to the hinge parts as a result of a LOCA.

The jackscrew locking devices prevent the jackscrews from turning during service. The replacement locking devices have the same function as the originally installed locking devices, which is to prevent the jackscrew from turning during operation.

Observable Effects:

A visual (VT-3) examination of 100% the accessible surfaces, including the overall valve symmetry in the mounting ring and the overall jackscrew thread extension from the lower retaining ring threaded flange is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-3 10-year ISi activities.

The vent valve locking devices are being examined to detect loss of material from the original locking device (wear associated with jack screw spring and pressure plate), cracking of the original locking devices (key ring and pin), and cracking of the modified locking devices, including detection of fractured or missing locking cups and welds or the bolted block.

The specific relevant conditions for the original locking devices are: evidence of damage or wear to the U-cover, misalignment of the pressure plate with the jackscrew, U-cover, or spring



• • • • • • • • • • • • • • • • • • • .I • • • •1 ., • • • ., • • • • • • • • • • •• • • •

• • • • • • • • • • • • • • • • :. • 1e

1: • 1.

• • • • • • • • • • • • • • • • • • • •


retainer, damage to the pressure plate or spring retainer, observations that the jackscrew is out of the design configuration (e.g., observation of a broken or cracked jackscrew, incorrect engagement of the jackscrews), and observations that the valve is not symmetrical in the mounting ring or the jackscrew thread extensions from the lower retaining ring threaded flange are unequal.

7

The specific relevant conditions for the modified locking devices are: evidence of fractured or missing locking cups and welds, a torn crimp or visual evidence that the crimp and sleeve slot are not aligned or engaged, a fractured or missing bolted block, missing cap screws associated with the crimped locking cups, observations that the jackscrew thread extensions from the lower retaining ring threaded flange are unequal, and observations that the valve is not symmetrical in the mounting ring .

Possible examination outcomes:

• Original locking devices o No relevant and/or reportable conditions identified o Observation that the vent valve is not symmetric in the mounting ring o Observation that the overall jackscrew thread extension from the lower retaining

ring is unequal and/or incorrect o Observation the jackscrew is out of design configuration o Observation of U-cover damage, wear, and/or incorrect positioning o Observation of pressure plate damage and/or misalignment

• Modified locking devices o No relevant and/or reportable conditions identified o Observation that the vent valve is not symmetric in the mounting ring o Observation that the overall jackscrew thread extension from the lower retaining

ring is unequal and/or incorrect . .

o Observation the jackscrew is out of design configuration o Observation ·of a cracked or fractured locking cup and/or locking cup weld o . Observation of a missing locking cup and/or locking cup weld o Observation of a cracked or fractured missing bolted block o Observation of a missing bolted block o Observation of a torn crimp or visual evidence that the crimp and sleeve slot are

not aligned or engaged o Observation of missing cap screws associated with the locking cups

Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the vent valve locking devices involves the following steps and inputs:

• Determine impact of observation(s) on future function of component item(s), weld(s), and overall vent valve function and/or design configuration




• Determine if the vent valve assembly can be left as-is (with observations) or if any actions (up to vent valve replacement) need to be taken for the vent valve (both component item and overall vent valve) to perform its intended function

Existing Documentation:

• Several previous analyses of observations have been performed with varying results (vent valve left as-is versus vent valve replacement)

• Bypass leakage calculations are available

• Exercise load calculations are available

8

• Original design calculations, such as vent valve opening force calculations, are available

What observations trigger examination into the Expansion category?

There are no expansion items for these component items.



• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •

• • • • • ···; • • •• ! •

• •• • :• i. • .j • • • • • • • • • • • • • • • • • • • • •• • • • • •


2.1.2 Vent Valve Bodies

Cast austenitic stainless steel (CASS) vent valve bodies are subject to thermal aging embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the material's reduced fracture toughness, such a condition could potentially lead to cracking. There is no known confirmed cracking of CASS material in PWR reactor vessel internals applications.


Vent valve assemblies are installed in the mounting rings in the core support shield. For all normal operating conditions, the vent valve is closed. In the event of a rupture of the reactor vessel inlet pipe, the valve will open to vent steam generated in the core directly to the break, thus permitting the core to be flooded and adequately cooled after emergency core coolant has been supplied to the reactor vessel. Each vent valve assembly consists of a hinged disc, a valve body with sealing surfaces, a split retaining ring and fasteners, and an alignment device (to maintain the correct orientation). Each vent valve assembly can be remotely handled as a unit for removal or installation. Vent valve component parts, including the disc, are designed to minimize the possibility of loss of parts to the reactor coolant system, and all operating fasteners include a positive locking device. The hingeddisc includes a device for remote testing and verification of proper disc function. The external side of the disc is contoured to absorb the impact load of the disc on the RV inside wall without transmitting excessive impact loads to the hinge parts as a result of a LOCA.

Observable Effects:

A visual (VT-3) examination of accessible surfaces of 100% of the vent valve bodies on the plenum side (core side) of the vent valve is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-310-year ISi activities unless an applicant/licensee provides an evaluation for NRC staff approval that justifies a longer interval between inspections ..

The vent valve bodies are being examined to detect cracking; this includes the detection of fractured vent valve bodies or surface irregularities (e.g., damaged, grossly cracked, or missing portions) associated with the vent valve bodies .


• No relevant and/or reportable conditions identified

• Observation that the vent valve is not symmetric in the mounting ring

• Observation of a vent valve body incorrectly positioned

• Observation of a vent valve body with surface irregularities (e.g., damaged, grossly cracked, or missing portions)

• Observation of a fractured vent valve body

9


••• )"his record was final approved on 3/13/2018 2:13:14 PM. ( This statement was added by the PRIME system upon its validation)



The general analytical methodology to be used for acceptance criteria for the vent valve bodies involves the following steps and inputs:

• Determine impact of observation(s) on future function of vent valve body and overall vent valve function and/or design configuration

10

• Determine if the vent valve assembly can be left as-is (with observations) or if any actions (up to vent valve replacement) need to be taken for the vent valve (both vent valve body and overall vent valve) to perform its intended function


• Calculations of ferrite content have been performed for various vent valve bodies

• A statistical assessment of PWR RV internals CASS materials has been made for the PWROG (PWROG-15032-P, Rev. 0-A [13]).

• Several previous analyses of observations have been performed with varying results (vent valve left as-is versus vent valve replacement)

• Bypass leakage calculations are available • Exercise load calculations are available • Original design calculations, such as vent valve opening force calculations, are

available



• • • • • • • • • • • • • • • • • • • ., • • • • • • • • , . • • • . ) • • • . , •• • •• • • •

• • • • • ,. • • • • • ., • •• • • • • • • • • • ,. • • • • :e i. :e • • • • ,. ,. i., • • • • •


2.1.3 Lower Grid Rib Section

The lower grid rib section is subject to irradiation embrittlement, which if a flaw would be present and it is subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is no known confirmed cracking of lower grid rib section material in PWR reactor vessel internals applications .


The lower grid assembly, which includes the lower grid rib section, provides alignment and support for the fuel assemblies, supports the core barrel assembly and flow distributor, and aligns the incore monitoring instrumentation (IMI) guide tubes with the fuel assembly instrument tubes. The lower grid rib section is a disk through which 177 squares are machined out, leaving a grid with "ribs." The square holes align with the fuel assembly locations in the core. There are additional holes about the periphery of the disk to permit a small bypass flow of reactor coolant up behind the baffle plates in the core barrel.

Observable Effects:

A visual (VT-3) examination of 100% of the accessible surfaces of the lower grid rib section heat-affected zone (HAZ) adjacent to the spider-to-lower grid rib section welds is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-3 10-year ISi activities unless an applicant/licensee provides an evaluation for NRC staff approval that justifies a longer interval between inspections .

If relevant flaw(s) are identified, follow-on examination by VT-1, ET, or UT may need to be performed to confirm and characterize the length or both length and depth of the observation(s).

The lower grid rib section is being examined to detect cracking, including the detection of readily detectable cracking .


• No relevant and/or reportable conditions identified • One or more lower grid rib section HAZ location(s) adjacent to the spider-to".'lower grid rib

section welds are identified with one or more minor (short) crack-like indicator(s) • One or more lower grid rib section HAZ location(s) adjacent to the spider-to-lower

grid rib section welds are identified with one or more large (long) crack-like indication(s)

• One or more lower grid rib section HAZ location(s) adjacent to the spider-to-lower grid rib section welds are identified with missing material


The general analytical methodology to be used for acceptance criteria for the lower grid rib section involves the following steps and inputs:


*** This record was final approved on 3/13/2018 2:13:14 PM, ( This statement was added by the PRIME system upon its validation)


• Record length or length and depth of crack-like indication

• Perform appropriate fracture mechanics analysis

• Determine future inspection interval based on observed condition


Minimal documentation currently exists for the lower grid rib section.

• A lower bound fracture toughness value for irradiated austenitic stainless steels has been documented (see MRP-175, Revision 1 [14])

• Crack growth rates from EPRI were issued in 2014 (See EPRI TR 3002003103 [9])

12



• • • ., • • • • • •· • ,. • • • • • • • . , • • • • • • • • • • • • • • • • • l •

• • . , • •

• • • • • • • • • • • • • . , • • • • ·• • • • • • ' ' :• • • • . , • • • • • • . , • • • • •


2.2 COMBUSTION ENGINEERING PLANT DESIGN

2.2.1 Cross Reference Matrix

A cross reference matrix is provided to identify the methodology to be applied for each component as well as any applicable RAls issued by the NRC during the course of their review of MRP-227, Rev. 1. Table 1 provides the cross reference matrix for Combustion Engineering components .




Table 1. Combustion Engineering Acceptance Criteria Methodology and RAI Cross Reference Matrix

Component 10<1> Component Name<1

> Methodology121

C1 Core Shroud Assembly (Bolted) - Core Shroud Bolts CE-ID: 1

C1.1 Core Shroud Assembly (Bolted) - Core Support Column Bolts CE-ID: 1.2

C1.2 Core Shroud Assembly (Bolted) - Barrel-Shroud Bolts CE-ID: 1.1

C2 Core Shroud Assembly (Welded) - Core Shroud Plate-Former Plate Weld CE-ID: 2 & Interim Guidance in 2.2.2

C2.1 Core Shroud Assembly (Welded) - Remaining Axial Welds CE-ID: 2.1 & Interim Guidance in 2.2.2

C3 Core Shroud Assembly (Welded) - Shroud Plates CE-ID: 3 & Interim Guidance in 2.2.2

C3.1 Core Shroud Assembly (Welded) - Remaining Axial Welds CE-ID: 3.1 & Interim Guidance in 2.2.2

C3.2 Core Shroud Assembly (Welded) - Ribs and Rings CE-ID: 3.1

C4 Core Shroud Assembly (Bolted) -Assembly CE-ID: 4

C4a Core Shroud Assembly (Welded) -Assembly CE-ID: 5

C5 Core Support Barrel Assembly...;.. Upper Flange Weld CE-ID: 6

C5.1 Core Support Barrel Assembly - Lower Girth Weld CE-ID: 6.1

C5.2 Core Support Barrel Assembly - Upper Girth Weld CE-ID: 6.2

C5.3 Core Support Barrel Assembly - Upper Axial Weld CE-ID: 7.1<4> & Int. Guidance in 2.2.2

N/A<3> Core Support Barrel Assembly- Upper Core Barrel Flange Evaluation Not Required

C5.4 Lower Support Structure - Lower Core Support Beams CE-ID: 6.4

C6 Core Support Barrel Assembly - Middle Girth Weld CE-ID: 7 & Interim Guidance in 2.2.2

C6.1 Core Support Barrel Assembly - Middle Axial Weld CE-ID: 7.1 & Interim Guidance in 2.2.2

C6.2 Core Support Barrel Assembly - Lower Axial Weld CE-ID: 7.1 & Interim Guidance in 2.2.2

C7 Core Support Barrel Assembly - CSB Flexure Weld CE-ID: 9

ca Lower Support Structure - Core Support Columns CE-ID: 8

C9 Lower Support Structure - Core Support Plate CE-ID: 10

C10 Upper Internals Assembly - Fuel Alignment Plate CE-ID: 11

C11 Control Element Assembly -Instrument Guide Tubes CE-ID: 12

C11.1 Control Element Assembly- Remaining Instrument Guide Tubes CE-ID: 12.1

C12 Lower Support Structure - Deep Beanis CE-ID: 13 Notes:

1. The "Component ID" and "Component Name" are consistent with MRP-227, Rev. 1 [1 ]. 2. The "Component ID" identified for the Methodology is from WCAP-17096-NP-A, Rev. 2 [2] unless otherwise specified. 3. This component is no longer identified as Primary or Expansion in MRP-227, Rev. 1 [1]. 4. The CE-ID: 7.1 methodology is appropriate although this component previously corresponded to the Upper Cylinder (Including Welds) (i.e. CE-ID: 6.2) in MRP-227-A.

PWROG-17071-NP

14

Applicable RAIS [3]

N/A N/A N/A RAl20

RAls 12, 20

RAl20

RAls 12, 20

RAls 12

N/A N/A RAls 5, 20

RAl20

RAl20

RAl20

N/A RAls 10, 20, 24

RAls 5, 20

RAl20

RAl20

RAI 16

RAl9

RAI 16

RAI 16

N/A N/A RAI 10


••••.•.••••• Tare•d•W,atio'aoW1W11i\W~(--e·t·a-d,th-R·s··o·-a·)··. ,., •.••• ( •• I

• • • • • • • • • • •• • • ., • • • • . : •· • • • •• • • • • • • . ) • • • • • • •· • • •· • •



Current Requirement

Inputs and Assumptions

Modified Requirement


PWROG-17071-NP

o For weld locations subjected to fluence levels above 3xl021 n/cm2

(E> lMeV), a CGR model appropriate for the fluence level and material must be used and a justification for its use provided. Such models are available from sources like proprietary EPRl technical reports.

(Note that the remaining Inputs and Assumptions are unchanged) .

o For weld locations subjected to fluence levels above 3xl021 n/cm2

(E> lMe V), a CGR model appropriate for the fluence level and material must be used and a justification for its use provided. One appropriate model is available in EPRl Report Models of Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Light Water Reactor Environments, Volume 2: Disposition Curves Application [9]. ] .




2.3 WESTINGHOUSE PLANT DESIGN

2.3.1 Cross Reference Matrix

A cross reference matrix is provided to identify the methodology to be applied for each component as well as any applicable RAls issued by the NRC over the course of their review of MRP-227, Rev. 1. Table 2 provides the cross reference matrix for Westinghouse components.



• • • • • • • • • • • • • • • • • • • el • •·: •• • • • • •• • • : . • • • • • • • • • : . • •

••••••••••••••••••••••••••••••••••••••••••• WESTINGHOUSE NON-PROPRIETARY CLASS 3

Table 2. Westinghouse Acceptance Criteria Methodology and RAI Cross Reference Matrix

Component 10<1> Component Name111 Methodology121

..

W1 Control Rod Guide Tube Assembly - Guide Plates (Cards) W-ID: 1 & Interim Guidancel3J

W2 Control Rod Guide Tube Assembly - Lower Flange Welds W-ID:2

W2.1 Control Rod Guide Tube Assembly -

W-ID:2 Remaining CRGT Lower Flange Welds .

W2.2 Bottom-Mounted Instrumentation (BMI) System - BMI Column Bodies W-ID: 2.4 W3 Core Barrel Assembly - Upper FlangeWeld W-ID: 3 W3.1 Core Barrel Assemblv - Uooer Girth Weld W-ID: 4 & Interim Guidance in 2.3.4 W3.2 Core Barrel Assembly - Uooer Axial Weld W-ID: 4.1 & Interim Guidance in 2.3.4 W3.3 Core Barrel Assembly - Lower Flange Weld W-ID: 5 W3.4 Lower Internals Assembly - Lower Suooort Fon::iinq or Castinq W-ID: 2.2 N/A\41 Core Barrel Assembly - Core Barrel Outlet Nozzle Welds Evaluation Not Required W4 Core Barrel Assembly -Lower Girth Weld W-ID: 4 & Interim Guidance in 2.3.4 W4.1 Upper Internals Assembly- Uooer Core Plate W-ID: 2.1 W4.2 Core Barrel Assembly - Middle Axial Welds W-ID: 4.1 W4.3 Core Barrel Assembly - LowerAxial Welds W-ID: 4.1

W4.4 Lower Support Assembly - Lower Support Column Bodies

W-ID: 2.3 (cast) and W-ID: 3.2 (non-cast) (Both Cast and Non-Cast) ·

W5 Baffle-Former Assembly - Baffle-Edqe Bolts W-ID6 W6 Baffle-Former Assembly- Baffle-FormerBolts See Interim Guidance in 2.3.2 W6.1 Core Barrel Assembly - Barrel-Former Bolts W-ID 7.1 W6.2 Lower Suooort Assembly - Lower Support Column Bolts W-ID7.2 W7 Baffle-Former Assembly - Assemblv W-ID: 8 W8 Aliqnment and lnterfacinq Comoonents - Internals Hold-Down Sorinq W-ID: 9 W9 Thermal Shield Assembly - Thermal Shield Flexures W-ID: 10 & Interim Guidance in 2.3.3 Notes:

1. The "Component ID" and "Component Name" are consistent with MRP-227, Rev. 1 [1]. 2. The "Component ID" identified for the Methodology is from WCAP-17096-NP-A, Rev. 2 [2] unless otherwise specified. 3. Interim guidance was issued in OG-18-46 [15], which will be included in WCAP-17451-P. 4. This component is no longer identified as Primary or Expansion in MRP-227, Rev. 1 [1].

PWROG-17071-NP


17

Applicable RAls [3]

RAI 19 RAl20

RAl20

N/A RAls 5, 20 RAIS 5, 20, 26 RAls 20, 26 RAls 20, 26 RAI 14 N/A RAls 5, 7, 20 RAI 14 RAl20 RAl20

RAl9

N/A RAIS N/A N/A N/A N/A N/A



2.3.2 Baffle-former Bolts

W-ID: 7 Baffle-Former Assembly - Baffle-former Bolts [21

Current Requirement

Methodology

Goal: Must demonstrate that projected number of additional bolt failures will not threaten acceptable pattern prior to next scheduled inspection.

Data Requirements: Loads Bolting patterns Baffle design Fast neutron (dpa) distribution in the baffle-barrel region (if needed) Projected bolt failure rate

Analysis: Procedures for establishing acceptable bolting patterns for the barrel-to-former bolts have been established in [6] and [7]. This methodology has been reviewed and accepted by the NRC in a Safety Evaluation issued in 1998 (TAC No. MAl 152).

Acceptance Criteria: 1. Observed pattern ofunfailed bolts meets pre-defined acceptance criteria or is analyzed in real time. 2. Less than 50% of initial margin consumed.

Nr< (N - Nreq)/2 (See below) The observed pattern of failed bolts must be shown to meet the conditions for being an acceptable bolt pattern per the NRC-approved methodology of [13] and [14] and have a reasonable margin to protect against additional failures during the inspection interval. The margin is defined in terms of the number of intact bolts beyond the number required for the acceptable bolting pattern. The margin (M) at any time is simply:

M = N - Nreq - Nf

where

N = total number of baffle-former bolts Nreq = number of baffle-former bolts in acceptable pattern

. Nf= number of failed bolts

Assuming that there are no failed bolts at the beginning of life, the initial margin is simply: (N - Nreq). Nreq is determined via an Acceptable Bolting Analysis, using the NRC-approved methodology outlined in [13] and [14]. For operation through the next IO-year interval, require that no more than 50% of initial margin be consumed at the time of the inspection. Require that no more than 50% of initial margin be consumed for any subsequent inspection as well.

Approach: Generic work completed in previous PWROG program

18



• • • • • •· • • • • • • •-• • • • • • • • • • • • • • • • • • • • • • • •• • • • •• • ••

• • . , • • • • • • • •· • • • • • • • •• • • • ., • • • • • • •-• • • • • • • • • • • • •

---------------------------------------------------~



Methodology

Goal: Must demonstrate that bolting patterns provide sufficient margin to acceptance criteria for continued operation through the re-inspection interval.

Data Requirements: Loads

Analysis:

Plant Desie:n

Down-Flow

Upflowand Converted Upflow Notes:

UT inspection results Test data for empirical model (as necessary) Baffle-former assembly design information including any applicable modifications Fast neutron (dpa) distribution in the baffle-barrel region (ifneeded)

1. Structural integrity ofthe baffle-former assembly must be demonstrated by an evaluation of the bolting pattern. One approach to accomplish this is to perform an acceptable bolting pattern analysis (ABPA) per the methodology described in WCAP-15029-P-A [6], which has been reviewed and accepted by the NRC via Safety Evaluation .

The bolting pattern considers functional bolts to include those with satisfactory UT inspection results or replacement bolts, as well as any portion of untestable bolts determined to be acceptable per step 2. Otherwise, the un-inspectable bolts and those that show visual or UT indication of failure are treated as non-functional. This evaluation must consider any plant-specific variations of operating conditions from analysis assumptions, and may credit plant modifications such as upflow conversion to reduce bolt load .

2. The degradation is categorized as either "Typical" or "Accelerated" based upon plant design and the inspection results. The percentage of bolts with indications is based upon the fraction of the total quantity. This categorization considers the likelihood of additional bolt failures due to IASCC over a typical 10-year reinspection interval. The degradation category for the observed and assumed failed bolts shall be determined using the following table:

Inspection Results<2> · Dee:radation Catee:ory Less than 3% indications with.no clusteringl11 Typical 3% or greater indications cir clustering<1l Accelerated · Less than 5% indications with no clustering\!! Typical· 5% or great~r indications or clustering\!/ · Accelerated

1. Clustering is defined as three or more adjacent defective BFBs in the vertical or lateral directions on the same baffie plate, or more than 40% defective BFBs on the same baffie plate [4]. Untestable bolts should be reviewed on a plant-specific basis for determination if these should be credited as structural members when classifying clustering .

2. These criteria are considered applicable to subsequent inspections until an appropriate degradation rate threshold is established through operating experience .

3. Additional margin to account for potential degradation through the next reinspection interval must be provided. Plants that exhibit "Accelerated" degradation must apply the Probabilistic Model, but those with "Typical" degradation may apply either the Probabilistic Model or Margin Ratio method as defined in the Acceptance Criteria .

Note: Any plant-specific evaluation used to extend the re-inspection interval beyond those defined in MRP-2017-009 [4] is to be submitted to the NRC for information at least one year prior to the end of the current applicable interval for BFB subsequent examination [12] .

19




Acceptance Criteria: 1. The analysis requirements as outlined above are applied to ensure structural integrity of the bolting pattern.

2. The degradation category is established consistent with industry guidance.

3. The evaluation must include margin by applying one of the following methods:

Probabilistic Model (Typical or Accelerated Degradation)

A. An empirical model of the degradation effect is developed from plant-specific design information and test data. The model is calibrated to plant-specific bolt inspection results, or otherwise applies bolt inspection results from similar plants after accounting for relevant differences as appropriate.

B. Determine patterns of predicted failed bolts at the end of the re-inspection interval based upon the likelihood of failure. Justification shall be provided that an appropriate degradation rate has been applied. The evaluation is allowed to credit bolt replacements, or plant improvements such as upflow conversion.

C. A probabilistic reliability analysis is used to demonstrate reasonable assurance that the pattern of failed bolts will be acceptable at the end of the re-inspection interval. These techniques must consider both the total proportion of failed bolts and the spatial distribution, or pattern, of bolt failures. When predicting the total proportion of failed bolts, it is acceptable to consider results associated with the 95th fractile. For the potential spatial distributions of bolt failures, an evaluation showing acceptability of at least 95% of the predicted patterns at the end of the reinspection interval can be used. It is anticipated that this method will be used to demonstrate a IO-year re-inspection interval. Re-inspection intervals beyond IOyears may be justifiable in certain circumstances ( e.g., acceptable bolting pattern relies on only replacement bolts and the plant is in, or has converted to, an upflow configuration).

Margin Ratio (Typical Degradation Only)

Less than 50% of the initial margin may be consumed. Therefore, the quantity of bolt failures upon re-inspection must be less than or equal to the number of bolt

. failures that has occurred to date. ·

·. Nr < (N - N,eq) / 2

where

N = total number of baffle-former bolts Nreq = number of baffle-former bolts in acceptable pattern Nf= number of failed bolts

The margin (M) is:

M = N - Nreq - Nf

This degradation model may be applied at subsequent re-inspections, up to a IOyear re-inspection interval, when the "Typical" degradation category remains applicable.

20



• • •· •• • ., • • • • • • • • • ., • • • • • • • • • •-• • • • • • • • • • • • • • • • •

• • • • • • • I• :. 1., • •• • • • • !. I• I• • • • • ,. • • • • • • • 1•

i •

• • • • • • • • • •


2.3.3 Thermal Shield Flexure

W-ID: 10 Thermal Shield Assembly-Thermal Shield Flexures (21

Current Requirement

Methodology

Procedure: 5. Demonstrate that the usage factor for low cycle fatigue in the remaining supports, the top (blocks) and bottom (flexure) supports, including, flexures, blocks, pins, and bolts is less than 1.0 for the interval before the next inspection cycle. Stresses resulting from normal and upset load conditions, including cyclic thermal loads, as well as, seismic and LOCA conditions (LOCA loads may not be required for some plants) must be considered. Increased loading on the core barrel wall should also be addressed .

21

6. Remove additional flexures from the model and repeat steps 3 through 5 if the results of the frequency and usage factor evaluations are acceptable for the failure of one flexure.


Methodology

Procedure: (Note that no changes occurred to Steps 1 through 4) .

PWROG-17071-NP

5. Assess low cycle fatigue in the remaining supports, the top (blocks) and bottom (flexure) supports, including, flexures, blocks, pins, and bolts for the interval before the next inspection cycle. Stresses resulting from normal and upset load conditions, including cyclic thermal loads, as well as seismic and LOCA conditions (LOCA loads may not be required for some plants), must be considered .

6. Remove additional flexures from the model and repeat steps Jthrough 5 if the results of the frequency and fatigue assessment are acceptable for the failure of on·e flexure .

March 2018 Revision O




Current Requirement




o For weld locations subjected to fluence levels above 3xI021 n/cm2

(E> lMe V), a CGR model appropriate for the fluence level and material must be used and a justification for its use provided. Such models are available from sources like proprietary EPRl technical reports.

(Note that the remaining Inputs and Assumptions are unchanged).

o For weld locations subjected to fluence levels above 3xI021 n/cm2

(E> lMe V), a CGR model appropriate for the fluence level and material must be used and a justification for its use provided. One appropriate model is available in EPRl Report Models of Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Light Water Reactor Environments, Volume 2: Disposition Curves Application [9].


••• This record was final approved on 3/13/2018 2:13:14 PM. ( This statement was added by the PRIME system'upon its validation)

• • • • • • • • • • • • • • • • • • • • • • • • • •· • • •· • • •• • . , • • • • • • • • •

• • • • • • • • • • • ,.' , •. : -~

'

• • • • • • • • • • • • • • • • •• • • • • • • • • • • • • ,.


3 REFERENCES

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1). EPRI, Palo, Alto, CA: 2015. 3002005349 .

2. Westinghouse Document, WCAP-17096-NP-A, Rev 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements." August 31, 2016 (ADAMS Accession No. ML 16279A320)

3. U.S. Nuclear Regulatory Commission Letter, "Request for Additional Information for Electric Power Research Institute Topical Report MRP-227, Revision 1, 'Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline' (CAC No. MF7740)" May 15, 2017 (ADAMS Accession No. ML 17079A027) .

4. Letter from David Czufin and Brian Burgos to the MRP Integration Committee Members, Subject: Transmittal of NEl-03-08 "Needed" Interim Guidance Regarding Baffle Former Bolt Inspections for PWR Plants as Defined in Westinghouse NSAL 16-01 Rev. 1, MRP Letter 2017-009, March 15, 2017 (ADAMS Accession No. ML 17087A106) .

5. Letter from Bernie Rudell and Anna Demma to the MRP Members, Subject: Transmittal of NEI 03-08 "Needed" Interim Guidance Regarding Baffle Former Bolt Inspections for Tier 1 plants as Defined in Westinghouse NSAL 16-01, MRP Letter 2016-021, July 25, 2016 (ADAMS Accession No. ML 16211A054) .

6. Westinghouse Document, WCAP-15029-P-A, Rev. 1, "Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions," January 1999 .

7. Westinghouse Report, WCAP-15030-NP-A, Rev. 0, "Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions under Faulted Loaci Conditions," March 2, 1999 .

8. Westinghouse Nuclear Safety Advisory Letter, NSAL-17-1, Rev. 0, "Guide Tube Gu.ide Card Wear Attributed to Ion nitride Rod Cluster Control Assembly," January 16, 2017,

9. EPRI Report, Models of l"adiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Light Water Reactor Environments, Volume 2: Disposition Curves Application. EPRI, Palo Alto, CA: 2014. 3002003103 .

10. Westinghouse Document, WCAP-17451-P, Rev 1, "Reactor Internals Guide Tube Wear -Westinghouse Domestic Fleet Operational Projections," October 2013 .

11. Framatome Document, Framatome-18-00336, "PA-MSC-1567 Task 2: Framatome final inputs for MRP-227 RV Internals Inspection Acceptance Criteria Interim Guidance," February 8, 2018 .




12. Letter from Mike Hoehn II and Brian Burgos to the MRP Integration Committee Members, Subject: Transmit NRCD Technical Staff Assessment of NEI 03-08 Interim Guidance on Baffle-Bolting, MRP Letter 2017-035, December 12, 2017.

24

13. PWR Owner's Group Document, PWROG-15032-P, Rev. 0-A, "PA-MSC-1288 Statistical Assessment of PWR RVI CASS Materials," August 2015.

14. EPRI Report, Materials Reliability Program: PWR lntemals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision 1). EPRI, Palo Alto, CA: 2017. 3002010268.

15. Letter from Chris Wax to the MRP Members, Subject: Transmittal of Approved "Needed" Interim Guidance for Addressing Accelerated Guide Card Wear Issue Described in NSAL-17-1, (LTR-RIDA-17-270, Revision 0), PA-MSC-1471, PWROG Letter OG-18-46·, February 20, 2018.



• • • • • • • • • • • • • • • • • • • • • • • • •• • • • • • • • • • • • •• • • • • • •

---- - - ----- -- -----------------------------------~

• • • ••

.

• • ., • • • • • • • • • • • • • ,.

1• 1: ,. I ,. !e

• • • • • • • • ' • • • • • • •

PWROG-17071 Revision 0 Proprietary Class 3

**This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

ApprQval lnfQrr11ation ,. '•

;

Author Approval Radonovich David Mar-06-2018 16:44:23

Author Approval Carpenter Bradley T Mar-06-2018 16:57:05

Reviewer Approval Palamara Matthew J Mar-07-2018 14:48:09

Reviewer Approval Kielb John Mar-13-2018 13:19:16

Manager Approval Molkenthin James Mar-13-2018 13:26:56

Manager Approval Meikle Thomas L for Mcfadden John L Mar-13-2018 14:13:14

Files approved on Mar-13-2018


WJD1

Cross-Out

Documents

PWROG-17071-NP, 'WCAP-17096-NP-A Interim Guidance.'participating utility representatives of the Materials Committee. This document is a deliverable under PWROG Project Authorization