Materials APC (MAPC) · 9:00 am Lessons learned from MRP-227 –Discussion on Implications for EPRI Guidelines D. Czufin, TVA 9:30 am Summary of Topical Screening Process D. Czufin,

© 2016 Electric Power Research Institute, Inc. All rights reserved.

Scot Greenlee, Exelon, Technical ChairKurt Edsinger, EPRI

Wednesday, August 31, 2016

Materials APC (MAPC)

Date: August 30, 2016

2© 2016 Electric Power Research Institute, Inc. All rights reserved.

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Morning AgendaTime Topic Lead

8:00 am Welcome and Introductions, Agenda Review S. Greenlee, Exelon

8:10 am

MAPC Operations

Action Items

Taking a Step on “Ease of Doing Business”

Cross Comparison of Program Research Focus Areas (RFAs)

Other

K. Edsinger, EPRI

8:30 am Summary from Issue Program Self-Assessment D. Odell, Exelon

9:00 am Lessons learned from MRP-227 – Discussion on Implications for EPRI Guidelines D. Czufin, TVA

9:30 am Summary of Topical Screening ProcessD. Czufin, TVA

T. Hanley, Exelon

10:00 am Morning Break All

10:15 am Member Satisfaction Survey Kyle Rogers, EPRI

10:30 amRegulatory / Code Update

Update on Readiness for Second License Renewal (summary status from BWRVIP and PMMP meetings)R. Dyle, EPRI

11:00 am

BWRVIP EOC Report-out

Key Points from BWRVIP EOC Meeting

BWRVIP Program Summary (Key 2016/2017 Products & Challenges)

T. Hanley, Exelon D. Odell, Exelon

A. McGehee, EPRI

11:20 am

PMMP EC Report-out

Key Points from PMMP EC Meeting

MRP Program Summary (Key 2016/2017 Products & Challenges)

SGMP Program Summary (Key 2016/2017 Products & Challenges)

D. Czufin, TVAB. Rudell, ExelonA. Demma, EPRI J. Stevens, LuminantH. Cothron, EPRI

11:40 am PWROG MSC Program Summary (Key 2016/2017 Products & Challenges)H. Malikowski, Exelon

J. Molkenthin, PWROG

12:00 pm Lunch - Crescent City Ballroom (Mezzanine Level) All


Afternoon AgendaTime Topic Lead

1:00 pm

Progress on Knowledge Transfer / Training

Tribal Knowledge Approach

Recent examples from MRP

Concept from BWRVIP

D. Czufin, TVA

A. Demma, EPRI

A. McGehee, EPRI

1:45 pm Update from Institute of Nuclear Power Operations (INPO) C. Larsen, INPO

2:15 pm Chemistry Program Summary (Key 2016/2017 Products & Challenges)J. Goldstein, Entergy

D. Wells, EPRI

2:30 pm

PSCR Program Summary

Key 2016/2017 Products & Challenges

Strategy & Strategic Challenges

J. Cirilli, Exelon

TG Lian, EPRI

2:45 pm NDE Program Summary (Key 2016/2017 Products & Challenges)K. Hacker, Dominion

S. Swilley, EPRI

3:00 pm Afternoon Break All

3:15 pm

WRTC Program Summary

Key 2016/2017 Products & Challenges


D. Patten, FENOC

G. Frederick, EPRI

3:30 pm Summary of Demonstration Strategy for KOH Keith Fruzzetti, EPRI

3:45 pm

Environmentally Assisted Fatigue

Overall Approach

Analytical Effort and Regulatory Progress

Testing Plans

D. Steininger, EPRI

N. Palm, EPRI

4:30 pm

Wrap-up Develop Messages for NPC Executive Committee

Collect Action ItemsAll

4:45 pm Adjourn


Strategic Focus

Typical focus for January meeting is on financial performance and products delivered over the past year, as well as budgets proposed for year N+1. Typical focus for the August meeting is on overall strategic direction and emerging needs.


Materials Action Plan Committee (MAPC)

Source: EPRI Operations Protocol

“Beginning January 1, 2010, the EPRI Materials Degradation and Aging Action Plan Committee has the principal role for overseeing industry activities related to primary system materials and the continuing commitment to the Industry Materials Initiative (NEI 03-08).

It will accomplish this through a combination of direct governance over those key programs for which it is directly responsible and coordination with those working subject to the initiative, but not under the Materials APC.”


PMMP Executive Committee (EC)

Nuclear Power Council

Materials & Aging Action Plan Committee (MAPC)

Executive Chair:CNO Level Executive

Technical Chair: VP/Director Level Executive

BWR Vessel Internals Program (BWRVIP)

Materials Reliability Program (MRP)

Steam Generator Mgmt Project (SGMP)

Primary System Corrosion Research

(PSCR)

BWRVIP Executive Committee

NDE Action Plan Committee*

Other APCs not shown

PWR Owners Group Materials Committee (PWROG MC)**

PMMP (EOC)

Welding & Repair Technology Center

(WRTC)

BWRVIP EOC

Water Chemistry Action Plan Committee

Long Term Operations

*NDE APC coordinates with Materials APC and PWR Owners Group **Materials Committee has a representative on Materials APC

Materials Advisory Structure


Materials Advisory Structure (cont’d)

Action Plan Committee

Executive Chair: Fadi Diya, AmerenTechnical Chair: Scot Greenlee, Exelon

EPRI: Kurt Edsinger

BWRVIPBWR Vessel &

Internals Program

Exec ChairTim Hanley

Exelon

IC ChairDrew Odell

Exelon

Andy McGehee EPRI PM

MRPPWR Materials

Reliability Program

IC ChairBernie Rudell

Exelon

Anne Demma EPRI PM

WRTCWelding & Repair

Technology Center

Exec ChairBill Pitesa

Duke

IC ChairDan Patten

FENOC

Greg Frederick EPRI PM

SGMPSteam Generator Mgmt Program

IC ChairJim Stevens

Luminant

Helen CothronEPRI PM

PSCRPrimary Systems

Corrosion Research

IC ChairJim Cirilli, ExelonIan Armson, Rolls

Royce, Technical Chair

TG Lian EPRI PM

PWR Materials Management Program (PMMP)Exec Chair: David Czufin, TVA

EPRI PM: Robin Dyle

Exec SponsorTom McCaffrey

Entergy

Exec SponsorBrad Adams

Southern Nuclear

Exec SponsorRichard Bologna

First Energy


Other Attachments

NEI 03-08 Document List

Handling EPRI Proprietary Information

Anti-Trust Guidelines for EPRI Meetings and

Conferences

Minutes From Last Meeting

MAPC Committee List

Additional Program Materials – PSCR

Additional Program Materials – WRTC

Additional Program Materials – BWRVIP

Additional Program Materials – SGMP

Additional Program Materials – MRP

Additional Program Materials – PWROG MC

Additional Program Materials – Chemistry

Additional Program Materials – NDE

Other attachments are in a separate PDF file and can be downloaded from the Member Center


Together…Shaping the Future of Electricity


Kurt EdsingerDirector

Materials Degradation/Aging APC MeetingWednesday, August 31, 2016

MAPC Operations



Topic

Action ItemsTaking a Step on “Ease of Doing Business”Cross Comparison of Program Research Focus Areas

(RFAs)Other


Action Items


Action Items (1 of 3) – Info from July Coordination Call – Update with August

# Action Responsible Due Date Status / Summary of Closure

75

Issue programs to track identified gaps in training, include in research plans, and report to their executive committees annually on progress (i.e., as Graybook / strategic plans are endorsed). Coordination with the Advanced Training Initiative should be part of the overall approach where there is a synergy (Note: This is the remaining part of the gap for the Materials Programs from Action #41). Edsinger 1-Jul-16

Changed date to reflecting the estimated arrival of the "training matrix", which will organize the training that we have and help with the assessment of gaps / priorities.

67

Knowledge Transfer: Looking for something to capture / convey the "tribal knowledge". Description of final product being sought not clear / still under discussion. Action is to come back to MAPC with some suggestions or a recommendation on what should be considered or where we might start.

Czufin / Dyle 1-Sep-16

Czufin made a proposal at the Atlanta meeting that appeared to be widely supported. It was essentially to create program training using key EPRI and industry experts; video with voice-overs as appropriate to make it flow better and to fill any gaps identified when the package is assembled. The one to start with, in terms of applying the voice-over approach, might be the existing NEI 03-08 video for the training at Exelon.

78Develop screening process for regulatory approval of EPRI products

Czufin / Hanley 1-Sep-16

Currently drafting a screening process to determine which products need regulatory approval. This process would then be endorsed by NRC. 1st draft of a procedure was isssued this week for folks to review and comment on.

79Develop strategy for working with Code with anticipated decrease in engagement by NRC Dyle 1-Sep-16


Action Items (2 of 3)# Action Responsible Due Date Status / Summary of Closure

80

Define approach for improving ease of doing business (quicker to identify, quicker to address, simpler to implement through better use of technology, sharing more clearly, etc.). Edsinger 1-Sep-16

81

Make meaningful progress on use of "one-pagers" (overviews of the key information for any given subject) and leverage graphic where possible. Edsinger 1-Sep-16

Coming from the "International Perspectives" session and follow-on discussion

82

SGMP to respond to the foreign material issue described by INPO (eliminating/removing loose parts versus analyzing why they can be left in place). Cothron 1-Sep-16

Jim Benson will give a presentation at the PMMP on the foreign object project and this will be summarized as part of the SGMP report out to MAPC.

83

BWRVIP to respond to the INPO challenge regarding dry tube replacement (why it is a business decision and not a safety decision). McGehee 1-Sep-16

Plan to close action #83 during the BWRVIP report out at MAPC.

85Discuss status of xLPR applications group on MAPC calls to ensure there are a sufficient number of participants. Demma 1-Sep-16

59

SLR/LTO: We know where the technical gaps are (there are not many), but need a plan on what products need to be updated for SLR that can be referenced in the GALL revision process. The plan should assess which products are OK as-is, which are already expected to be revised in time for SLR and which need to be expedited. For the latter, they need to be incorporated into the plans of the Issue Programs. Provide update to MAPC ~ every 2 months. Reach out to Bernhoft / Gallagher to ensure coordination. (Note: #20 from Chicago Meeting minutes.) Dyle

extended to 31-Dec-16

Roadmap has been developed and provide to LTO for inclusion with the overall roadmap for SLR. Also planning to provide an update on SLR at the August MAPC meeting.


Action Items (3 of 3)

# Action Responsible Due Date Status / Summary of Closure

84Chemistry to provide update on the PAA/steam project at the next MAPC Meeting Fruzzetti 1-Feb-17

Team ran into some complications in the testing which forced testing to be halt while to assessing the complications / determining a revised path. Shifted action out by 6 months to provide update at January meeting.

62

Question related to BTP 5-3: How do owners manage loss of margin related to the issue, even if the license is not impacted? (Will be able to better address after the NRC reviews & accepts / comments on MRP-401/BWRVIP-287.) Hardin 15-Feb-17

NRC position still pending (no answer at January meeting). Note that if the EPRI position is accepted, then the loss of margin has been managed. The industry (PWRPG & MRP) effort to transition to use of direct fracture toughness (Master Curve) for RPV integrity evaluation will also recoup any margin lost from the BTP issue, but that is a longer term (~8 year) effort. Revised due date to reflect the fact that NRC told us on the April quarterly call that they are doing PFM work to evaluate the industry reports and would finish near the end of the year.


Ease of Doing Business


Taking a Step on “Ease of Doing Business” It was suggested from the International Session that one page

descriptions of key topic areas would be helpful Examples could include RPV, PWR Internals, BWR Internals, Jet

Pumps, Alloy 600, Foreign Objects, Primary Chemistry, etc. The one-pager could adopt a format like the following:

i. Describe the issueii. Summarize the key gaps, focus areas of current researchiii. List the guidance documents availableiv. List key technical references availablev. List the training availablevi. Provide a contact name for further inquiries

The one pagers could then go on the cockpits and act as a launch points for the topics


Example One-Pager Needed strategy for managing degradation on the

secondary side of steam generator tubing– Efficient method for foreign object inspections– Quantitative method for prioritizing foreign object

removal efforts– Eliminating buildup of deposits that can

accelerate corrosion, reduce heat transfer efficiency, disrupt thermal hydraulic conditions

ECT Technique Qualifications for Foreign Object Inspection

epriq.com

Foreign Object Removal Prioritization

10190391020989

SG Integrity Assessment Guidelines

3002007571

Foreign Object Handbook 3002007578Causes and Mitigation of TTS Denting

3002002197

Characterization of SG Deposits 3002002794Deposit Removal Strategies 3002005090Dispersant ApplicationSourcebook

1025317

PWR Secondary Side Water Chemistry Guidelines

1016555

Sludge Management Workshop 3002002756Secondary Side ManagementConference

Every two years

EPRI Technical Contacts: Brent Capell, James Benson


Research Focus Areas


Cross Comparison of Program Research Focus Areas

There are benefits of having a consistent naming and numbering convention for RFAs

Discussed on several occasions, including the Self-Assessment

It’s possible and probably the right thing to do


Research Focus Areas (RFA) Proposal

Use unique RFA # index for EPRI Materials R&D #RFA.#PROJECT.#TASK

Limit number of RFA’s (perhaps 15-20 across entire group)RFA’s can incorporate all key elements of “Roadmaps”

– Ensure that all EPRI R&D programs (i.e., internal to EPRI ) collaborations are incorporated into RFA’s

– Ensure that collaborations external to EPRI are incorporated into RFA’s

Ultimately need the Programs to work together and identify the optimal approach


Research Focus Areas (possible approach)

RFA DESCRIPTION 41.01.01 - PSCR 41.01.02 - SGMP 41.01.03 - BWR 41.01.04 - MRP 41.01.05 - WRTC

1 Reactor Internals Management

BWR Inspection & , RCS Piping and Components Inspection Technologies, NDE Technology

Reactor Vessel Internals Assessment, Modeling and Inspection

2 Irradiated Materials

IASCC mechanistic models / correlations, IASCC mechanism, De-tensioning of X-750 annulus spacers as a result of irradiation-induced creep and void swelling, The effects of irradiation aging or synergetic aging on fracture toughness

Thermal and Irradiation Effects on Stainless Steel Reactor Internals

Reactor Vessel Internals Irradiated Materials Testing

3 Nickel Base Materials

PWSCC degradation in Alloy 600/82/182 components (especially DMW)

Managing Corrosion Mechanisms Affecting SGs, Overall significance of ODSCC and PbSCC on steam generator tubes, Sherlock

Alloy 600/690 Management, Mitigation, and Inspection

4 Probabilistic Analysis Methods LBB (xLPR)

5 Stainless Steel Materials

Degradation mechanism for Ti-stabilized stainless steels, Factors that lead to the cracking of VVER steam generator collect or welds

BWR-Related Corrosion Research

Stainless Steel Degradation Mechanism Studies, Reliability of CASS Pressure Boundary Components

6 Thermal and Vibration Fatigue Fatigue Management (Dryer) Fatigue Management (Thermal

and Vibration Fatigue)

7 Replacement Materials

Development of irradiation resistant alloys and IASCC mitigation strategy

Management of High Strength Alloys

Replacement Materials Testing (Alloy 690/52/152)

8Pressure Vessel and Low Alloy Steel Materials

RPV embrittlement, The vulnerability of SCC crack propagation in low alloy steels

Reactor Pressure Vessel Integrity

Assessment for RPV Integrity, Boric Acid Corrosion


Research Focus Areas (possible approach)RFA DESCRIPTION 41.01.01 - PSCR 41.01.02 - SGMP 41.01.03 - BWR 41.01.04 - MRP 41.01.05 - WRTC

9 Environmentally Assisted Fatigue EAF mechanisms EAF EAF

10 Wear Understanding Wear Mechanisms in SG Tubing

11 MitigationEffectiveness of HWC mitigation to IASCC Mitigation of BWR Material

Degradation Peening of 600/82/182 Thermal spray, Coatings, and Hardfacing Applications

12 Weld Repair Welding of irradiated materials+D9

BWR Irradiated Materials Welding

Irradiated Materials Welding Solutions, Small Bore Piping Asset Repair Solutions, Buried Pipe Asset Management / Repair Solutions, Repair Solutions for Structures: Containment, Fuel Pool Asset Management, Spent Fuel Storage

13 Welding Fabrication

Optimized Joining, Fabrication, and Repair Processes (including WRS), Identify, Research, Develop, and Mature Advanced Welding Processes Development

14 Optimized Weld Materials

Nickel-Base Filler Metal Weldability and New Alloy development

15 Flow PhenomenaMechanism of cracking in CANDU primary heat transfer (PHT) feeder pipe bend

Development of State of the Art Software for Analyzing Steam Generator Flow Conditions

Management of Jet Pump Flow-Induced Vibration CFD for thermal fatigue

16 Steam Generator Inspection

Development and Employment of Eddy Current Technology Improvements, Foreign Object Inspection Technologies, Sherlock


Other


Expanded Mission and Scope for PSCR

This program will have the overall responsibility for materials testing, characterization, and modeling within the Materials Programs. It will directly support the other Programs and its scope will be completely integrated with their strategic plans. The Program will also support other areas of EPRI with materials expertise (e.g., FRP, Used Fuel, etc.), and will act as a focal point for materials-related TI scope. Additionally, the program will provide the key engagement point for Base members that do not participate in the other programs (e.g., those with other reactor designs).


PSCR Leadership

TAC Chair: Jim Cirilli, Exelon – Vice Chair: open

Technical Chair: Ian Armson, Rolls RoyceEPRI Program Manager: Anne DemmaTG Lian assigned to support EPRI engagement in Europe


Materials Organizational Chart – August 2016

Al AhluwaliaKyle AmbergeBrian BurgosPaul CrookerTim HardinCraig HarringtonMike McDevitt

Peter ChouGabriel IlevbareRaj PathaniaCem TopbasiJean SmithDavid Steininger

Jim BensonBrent CapellRich GuillSean KilRick Williams

Bob CarterJohn HoslerWayne LuncefordAmardeep MehatNathan PalmChuck Wirtz

Dana CouchSteve McCrackenNick MohrArtie PetersonBen SuttonJon TatmanStacey Wells

Kurt EdsingerMaterials

OpenMRP

Helen CothronSGMP

Anne DemmaPSCR

Andy McGeheeBWRVIP

Greg FrederickWRTC

Anna DeshonRobin DyleDavid GandyJennifer MaChuck Welty




Drew OdellExelon – BWRVIP Integration Chair

Nuclear Power Council Advisory MeetingWednesday, August 31, 2016

NEI 03-08 - 2016 Self Assessment Report



NEI 03-08 - 2016 Self Assessment Report

NEI 03-08 requires Materials Issue Programs to perform a self assessment every 3 years.

Self Assessment performed week of July 25th in Charlotte.All representatives were present including: EPRI Materials

Director, EPRI Project Managers, Issue Program IC Chairs, INPO Asset Management representative, Duke Energy Utility Participant and various EPRI personnel.

IP’s in attendance were: BWRVIP, Chemistry, MRP and SGMP.

All IPs were found to be meeting the intent and main objectives of NEI 03-08.


NEI 03-08 - 2016 Self Assessment Recommendations

A revision to NEI 03-08 is needed to align with how the Issue Programs (IPs) have evolved.

IPs should fully assess the implications of all materials related “safety communications” from OEMs. (Tech Bulletins, NSAL, SC, SIL etc.)

IPs should ensure “issues” have visibility in the Issue Management Tables (IMTs) or IP project lists.

IPs (members and vendors) improve flow of restricted or proprietary information to ensure analysis includes a complete picture. (e.g. – RCA from D.C. Cook, subsequent analyses and potential impact from Cook)


NEI 03-08 - 2016 Self Assessment Recommendations (cont.)

IPs consider how trending of inspection data or OE could provide early indicators.

IPs consider collecting data from limiting plants. IPs could sponsor industry surveillance programs to collect and share data.

Improve IP administrative procedures– Document a common Emergent Issues Protocol– Document Research Focus Areas– Consider a more formal OE determination process of “Emergent

Issues”


NEI 03-08 - 2016 Self Assessment Next Steps

Complete Self Assessment report – September 2016. Obtain reviews, concurrence and approval of Self

Assessment report – October 2016. Enter all recommendations into EPRI Corrective Action

Process – October 2016.




Kurt EdsingerDirector, Materials

MAPC MeetingWednesday, August 31, 2016

MRP-227-ACorrective Action Report



Problem Statement

During the Spring 2016 outage season, two plants (Indian Point 2 and Salem 1) were performing inspections guided by MRP-227-A (“Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines”) and encountered a larger than expected number of baffle-to-former bolt indications and failures.

In hindsight, there was industry experience (most notably DC Cook 2 with 18 failures) that potentially could have been more fully leveraged as the guidance was developed and revised.

CAR opened – Unique for an EPRI CAR, but with potentially broad implications


Cause The cause of this issue was that the factors leading to bolt degradation were

not well enough understood to develop a strategy that would not appear overly conservative for the vast majority of plants and targeted guidance was not developed for plants that could have been identified as benefitting from additional conservatism.

Incomplete data, counting on forthcoming measurements to improve understanding

A Utility/Vendor root cause report that was not fully shared Under-appreciation of potential consequences / impact, especially related to

outage extensions and emergent resource needs Having only one US plant with issues and thinking they must be unique Believing it could not happen here (even though EDF saw it)

Contributing Factors


Immediate & Corrective Actions Evaluation

– Westinghouse Part 21 Evaluation (5/13/2016)– Westinghouse NSAL-16-1 (7/5/2016)– AREVA CUSTOMER SERVICE BULLETIN NO. 16-02 (7/14/2016)

Communication– Industry Communications

Indian Point 2 Emerging Issue Call (3/31/2016) Salem 1 Emerging Issue Call (5/10/2016) PMMP communication on Westinghouse TB12-5 (6/1/2016)

– NRC Communication Meeting with Bill Dean (5/25/2016)

– Congress Communication Meeting with staffers organized by NEI (6/6/2016)

Guidance– Interim Guidance issued as MRP 2016-021 (7/25/2016)


Proposed Preventive Actions1. Assess Safety Communications: IPs should assess the implications of all materials-

related safety communications (Tech Bulletins, NSAL, SILs, etc.) on current NEI 03-08-related guidance

– This is not routinely happening now (only happens when the communication is causing difficulty in the industry) and Westinghouse TB 12-5 was a missed opportunity. This should be similar to how OE / emerging issues are handled. (Note that, for this particular case, it would not have changed the outcome much since the inspection techniques were not yet qualified, but it could have enabled the utilities to be better prepared regarding assessment and replacement.)

2. Escalate IP Sharing Issues: IPs should improve sharing of Proprietary Information or ensure that the entity that controls the information is the one issuing the guidance

– The restricted flow of information was likely a contributing factor as not everyone was able to see the complete picture (root cause from DC Cook, subsequent analyses, and potential impact)

3. Identify Limiting Plants: IPs should identify (to the best of their ability) and communicate the limiting plants when guidance is issued (probably not directly in the guideline, but in some way)

– This could be helpful in cases where the guidance requirements bound a large fraction (or all) of the fleet.


Potential Preventive Actions (cont’d)4. Sponsor Surveillance Programs: IPs should consider sponsoring industry

surveillance programs to collect and share the data in cases where the industry would benefit from collecting the data (i.e., from limiting plants) and the plants do not have plans to perform those measurements in a timeframe desired by the fleet

5. Ensure Issues have Visibility: IPs should ensure "issues" have visibility in the Issue Management Tables (IMTs)

– In this case, there is no explicit IMT gap for Baffle Bolts, although there is a general IMT gap for irradiated materials testing (P-AS-14a and 14b)

6. Trend Operating Experience: IPs should trend OE and/or inspection results to look for early indicators

7. Leverage Review Visits: IPs should provide key topics to INPO (and others) to support planned review visits




David Czufin (TVA), PMMP Executive ChairTim Hanley (Exelon), BWRVIP Executive Chair


Summary of Topical Screening Process



Background This is an action item from the January 2016 MAPC meeting that later

became a Deliver on the Nuclear Promise (DNP) initiative EPRI NEI 03-08 issue programs have drafted a document screening

process to be used in conjunction with the existing topical report review and approval process currently used by BWRVIP and MRP

Analogous to 50.59 process Initial application focus on BWRVIP documents

– BWRVIP has majority of documents with implementation contingent on NRC approval

– There are clear examples of difficulty in obtaining NRC support in reasonable and timely reviews of optimized BWRVIP I&E guidance

Proposed process includes initial “applicability evaluation” and more detailed “screening evaluation”


Key Elements (1 of 2)Applicability Evaluation

– Used to determine when further document screening is applicable

– Document screening only applicable for reports containing EPRI IP controlled aging management guidance associated with prior NRC review and approval

– Excludes NRC review for documents not previously approved by NRC or without explicit commitments to NRC

– As drafted, does no harm to existing processes not currently dependent on topical report approval

(1c)Prior guidance approved by NRC via SE or is credited to meet a specific IP commitment to NRC?

(1a)Topical report revision / replacement or new topical report contains implementable aging management guidance?

(1d)Does the guidance represent a technical change to NEI 03-08 Mandatory or Needed elements?

(1e)Could implementation of the guidance result in a change to aging management elements other than the inclusion of new elements?

Perform screening evaluation of new / revised aging management guidance.[Go to Fig. 2, box (2a)]

YES

YES

No EPRI IP controlled implementation elements. Screening not applicable.

Document may be generically released by the IP for implementation without NRC approval.

NO

NO

YES

(1b)Implementable aging management guidance elements are applicable to safety-related SSCs or to nonsafety-related SSCs generically subject to aging management for license renewal?

YES

YES

Guidance relevant to asset preservation only and may be generically released for implementation without NRC approval.

NO

NO

NO

DRAFT “APPLICABILITY EVALUATION” PROCESS


Key Elements (2 of 2)Screening Evaluation would

allow implementation without NRC SE for:– Revisions consistent with NRC

precedent / prior NRC approval– Determinations based on

qualitative risk-based evaluation, using methods consistent with those previously approved by NRC

– Determinations based on quantitative risk-based analysis with results meeting criteria accepted by NRC

DRAFT “SCREENING EVALUATION” PROCESS

(2c) – Prior Acceptance of Similar Methodologies

Direct Precedent:Are the guidance changes consistent with either direct NRC precedent associated with analogous applications?

OR

Analysis Method Precedent:Are guidance changes directly supported through application of improved analytical methods that use inputs, evaluation methods, and margins on acceptance criteria essentially the same as those previously approved by NRC for similar applications?

NRC approval of change(s) required prior to generic release for implementation.

(2e)Are the evaluation results sensitive to new OE or R&D results?

Release change(s) for generic implementation without NRC approval.

(2a) – Component Safety Function(s)Is the proposed change predicated on conclusions based on revised component safety function or consequence of failure assessment(s)?

(2b) – Impact on Availability of Future OEDoes the proposed change recommend complete elimination of future examinations for any distinct component location previously subject to periodic inspection?

NO

(2d) – Change in Risk

Qualitative Risk Assessment:Are the changes directly and clearly supported by a generic qualitative assessment of risk (e.g., FMEA) that concludes the proposed changes are, as a minimum, risk-neutral?

OR

Quantitative Risk Assessment:Are the changes directly and clearly supported by generic risk-based analyses that demonstrate the proposed changes have a negligible impact on level of safety using bases previously accepted by NRC?

NO

NO

YES

NO

(2f)Can bases for reevaluation to address new data be established and included as a condition of release for implementation?

NO

YES

YES

NO

YES

YES

YES


Schedule, Status & Impact Schedule as defined by MAPC

– Aug 2016 – Development of proposed screening process (COMPLETE)– Aug 2016 – Executive leadership presents approach to NRC (COMPLETE)– Aug 2016 to May 2017 Executive leadership works to obtain NRC acceptance (in progress)– May 2017 – If process approved AND SE not received on optimized version of BWRVIP-41,

withdraw from NRC review and implement based on document screening process– Present – Feb 2017 MAPC meeting – Identify schedule for revision of cost beneficial NEI 03-08

documents (in progress) Feedback from NRC @ August technical exchange meeting is that NRC is attempting to

determine an appropriate process for review (also will need to assign a primary POC) BWRVIP impact

– Accelerated implementation of: BWRVIP-41 Rev. 4 (Optimized Jet Pump I&E Guidance) BWRVIP-76 Rev. 2 (Core Shroud I&E Guidance)

– Could also result in acceleration of I&E guidance optimization by the BWRVIP for other components (should a success path via screening be identified)

– Screening directly benefits U.S. utilities operating under NEI 03-08 initiative. However, concept and principles could be beneficial for international utilities in interacting with regulators




2016 Nuclear Sector Member Satisfaction



Member Satisfaction - Background

EPRI has captured member satisfaction feedback in various forms for many years

Current member satisfaction survey adopted by Board in 2006

Results reviewed regularly with Board– one of Corporate Performance Indices (CPIs)

Member feedback used to drive continuous improvement across EPRI

Helps prioritize efforts – focus on areas with greatest impact on

satisfaction


Nuclear Member Satisfaction Survey Results

Overall Performance


Technical Program Value

Overall Satisfaction

2015 Results

2010-2015 Trend

92.3%

Overall Performance


Technical Program Value

86.1%

93.2%

92.4%

75%

80%

85%

90%

95%

100%

2010 2011 2012 2013 2014 2015

• Impact of research on improving my business

• The program's strategic priorities and directions

• Quality of research results

• Relevance of research carried out by the program

• Technical staff expertise

Top ranked aspects of EPRI Experience

Who completed the Survey


Category Initiative Timeframe

Research and Development

• Research Focus Areas• Project Overview Forms• Quality Management Program Implemented 2016

Tech Transfer

• Executive Summary• Onsite EPRI updates/regional meetings• International workshops• International NPC• Digital Strategy (ongoing)

Implemented 2016

Simplification• On-line Pricing• Invoice Review• New Pricing Model

Implemented 2016

Website • New Search Engine• Member Center Improvements Implemented 2016

Improvement Initiatives

Listening and Responding to the Feedback of our Members


Digital Delivery Enhancements

Becomes


Becomes

Digital Delivery Enhancements


New Search EngineThe search engine gets smarter over

time based on use.

It tracks what people search and where they go with the results.

The more the search engine is used, the faster it learns.

As it learns, features such as relevance and search term recognition will

dramatically improve, and as a result improve your search experience.

You make the search engine better by using it!


Survey instrument

Key components …1. Who you are

without a name and organization, we can’t count your input!

2. Number of years you have been an Advisor

3. How we’re doing

4. How you assess EPRI value

5. Key improvement in ease of doing business

6. Value you have received from this Program


Survey instrument

Key components7. Rate each statement based on how

satisfied you are8. Rank the top 5 statements as

indicated in the instructions9. Would you recommend EPRI10.If you are not satisfied with us in any

area, please tell us why

9

10

7

8




Robin DyleSenior Technical Executive

Materials Degradation/Aging APC MeetingAugust 31, 2016

Regulatory / Code Updateon Readiness for Second

License Renewal



Topics

SLR issues– Actions/meetings to date– Successes– Items to appeal– Support for lead plants– VIP/MRP work

NRC– Tech mtg– Mgmt mtg

Code– Status of code cases– NRC changes


Overview of SLR

Effective January 2016 the SLR support work by BWRVIP and MRP became part of the routine program work.

Follow-up meetings held with lead plantsProject plans are doneWork is underway


EPRI Lead:Wayne Lunceford

[email protected]

BWRVIP Activities to Address SLR

mailto:[email protected]


Status & ScheduleWorking multiple technical areas in parallel to accelerate

completion schedule– Reactor internals fluence study– Generic BWR internals aging management review technical bases – RPV circumferential and axial weld probabilistic fracture mechanics

(PFM) evaluations– RPV equivalent margins analysis screening– EAF database development– 80-Yr Integrated Surveillance Program

Engaged with lead plant to ensure appropriate work prioritization(most recent meeting in July 2016)

Most of the technical work anticipated to be complete by the end of 2017


Reactor Internals Fluence Study Goals

– Provide comprehensive dataset of BWR/3-6 internals fluence

– Classify and categorize internal components according to material of construction and EOL fluence

– Identify any locations for which existing aging management guidance may not adequately consider / address EOL fluence for SLR

Approach– Create a consolidated fluence database from existing

data– Project service times when component locations may

exceed target fluence values Database will include all classes of U.S. BWRs

except:– BWR/3 with 724 assemblies (Dresden, Quad)– BWR/6 with 800 assemblies (Grand Gulf)

Results feed into updated AMR for BWR internals addressing LTO

Unit Type RTP (MWth)

Fuel Assy

RPV ID (in)

SHR OD (in) RPV Data Internals

Data

Monticello BWR/3 2004 484 205 167 X

Hope Creek BWR/4 3840 764 251 207 X X

Limerick 1 BWR/4 3515 764 251 207 X X

Limerick 2 BWR/4 3515 764 251 207 X X

Susquehanna 1 BWR/4 3952 764 251 207 X X

Susquehanna 2 BWR/4 3952 764 251 207 X X

Browns Ferry 2 BWR/4 3458 764 251 207 X

Cooper BWR/4 2419 548 218 177.5 X X

Duane Arnold BWR/4 1912 368 183 145 X X

FitzPatrick BWR/4 2536 560 218 177.5 X X

Hatch 1 BWR/4 2804 560 218 177.5 X X

Hatch 2 BWR/4 2804 560 218 177.5 X X

Peach Bottom 2 BWR/4 3951 764 251 207 X X

Peach Bottom 3 BWR/4 3951 764 251 207 X X

LaSalle 1 BWR/5 3546 764 251 207 X X

LaSalle 2 BWR/5 3546 764 251 207 X X

Clinton BWR/6 3473 624 218 185.5 X X

Perry BWR/6 3758 748 238 202 X X


BWR Internals Aging Management – Talking Points NRC license renewal staff have expressed concern about development of “80-year”

BWRVIP I&E Guidelines– Represents a fundamental disconnect in aging management approach– BWRVIP guidance applicable so long as conditions placed on the engineering evaluations

remain satisfied (e.g., accumulated neutron fluence, number of fatigue cycles)– Not tied to any single licensed operating period– Significant variation in plant design – some plants can apply “bounding” 60-year analyses for

80-years or more of operation AMR document addressing SLR for BWR internals being developed to address this

disconnect– Analogous to BWRVIP-74-A1 but addressing BWR internals LTO– Revisits technical basis evaluations that support inspection intervals to reevaluate suitability of

existing inspection intervals for operation beyond 60 years (e.g., generic flaw tolerance calculations)

– Addresses SLR SRP-SLR further evaluation items related to IASCC, Loss of Preload– Assessment consolidated in a single report in lieu of including “SLR evaluation” content

multiple inspection & evaluation guidelines

1 BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal (BWRVIP-74-A) 1008872

https://membercenter.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=000000000001008872


RPV Weld Probabilistic Fracture Mechanics (PFM) Issue:

– Many BWRs will need updated probabilistic analyses as a technical basis for RPV weld ISI exemptions and relief

– Recent work by U.S. NRC and ORNL indicated significant conservatism in BWRVIP-05 approach

– Use of the FAVOR Code can address limitations associated with both RPV circumferential and axial welds for initial LR and SLR (and beyond)

Status of BWRVIP Work:– Initial assessments performed by BWRVIP

using the currently available beta version of FAVOR indicate possible resolution path

– Work temporarily on hold pending issue of revised FAVOR Code by ORNL / NRC(revised code completion promised Sept 2016)

Adjusted Reference Temperature (F)

Con

ditio

nal P

roba

bilit

y of

Fai

lure

Source: PVP2015-45836, “Analysis of Circumferential Welds in BWRs for Life Beyond 60”


Screening to Assess RPV Equivalent Margins Analyses for SLR Generic EMAs provided for initial license

renewal in Appendix B of BWRVIP-74-A1

Many materials likely remain bounded by this generic EMA through SLR

Perform initial screening assessment to identify SLR needs– Project reductions in USE values for all available

plate, and weld materials using 80-year fluence estimates

– Identify those plants that are not projected to remain acceptable for SLR

For plants not projected to pass the initial screen, perform a second screen using method proposed by NRC as a replacement to the 10 CFR 50 Appendix G 50 ft-lb acceptance criteria SMAW Materials (BWRVIP-74-A, Fig. B-5)

1 BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal (BWRVIP-74-A) 1008872



Screening to Assess Environmentally-Assisted Fatigue (EAF) Margins Screening approach similar to that taken for RPV EMA being taken for

EAF– Anticipated that a majority of BWR component locations either can pass for

SLR using existing analysis or with additional analyses using accepted methods

Database of U.S. BWR EAF status under development– Results will identify scope of component locations requiring alternative / more

advanced analyses, as well as the magnitude of the issue for SLR– Preliminary results indicate that only a small percentage of BWR locations

exceed CUFen of 1.0 with environmental effects for 80-year operation Results should identify any substantial difficulties anticipated for BWRs

with regard to EAF TLAAs Data obtained will be applied to inform future EAF work


80-Yr Integrated Surveillance Program Development Status Overview Initial assessment of options complete

– Some data applicable for 80 years may be available from: 4th or reconstituted capsules in ISP host plants High lead factor SSP capsules

– No single option is most favorable in all areas of feasibility– Optimal approach would maintain the critical elements of the ISP to the greatest

extent possible, provide additional higher fluence data for the entire set of ISP materials, and be characterized as an extension to the existing ISP

Feasibility evaluation report published as BWRVIP-2951

Work in progress to gather data and perform investigations recommended in BWRVIP-2951

Draft report providing results of investigations will be completed in late 2016 Meeting with NRC targeted for Late 2016 to provide briefing on planned

80-year approach and obtain feedback1 Boiling Water Reactor Vessel and Internals Project, Reactor Vessel Material Surveillance for U.S. BWRs During the Second License Renewal: Feasibility Evaluation (BWRVIP-295) 3002007041



EPRI Lead:Kyle Amberge

[email protected]

MRP Activities to Address SLR



Lead Plant Activity Update PWR Component List

• Surry implement MRP-227-A1 A/LAI 1 & 2 and update lead plant RVI component list as necessary for SLR• This was not done for 1st LR since Surry was pre-GALL

• Confirm no Surry-specific component replacements or plant modifications since first license renewal appl.

• Provide Surry input to industry’s MRP-1912 component and material update for SLR applicability evaluations

1 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) 10228632 Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191) 1013234

http://membercenter.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=000000000001022863



MRP Update Mechanism Screening Criteria

• Simultaneous activity with lead plant RVI component list update actions

• Review MRP-1751 materials aging screening criteria and identify criteria that could change• Update of screening criteria as necessary• Included as part of planned MRP-1912 update

• NRC Staff briefing and follow-up technical discussions planned for first half of 2017

1 Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175) 10120812 Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191) 1013234




Joint (MRP and Surry) Component Screening / Evaluation

• Based on lead PWR RVI updated component list and MRP updated screening criteria

• Identify components with “new/changed” degradation mechanism

• Identify any previously identified “non-susceptible” component where extended operations would cause aging effect to become credible

• Provide these components for FMECA evaluation• Provide updated list of susceptible components for update of

MRP-1911 list of applicable components1 Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191) 1013234



Joint Component Screening and Evaluation (Cont.)• Develop re-screening and FMECA analysis of susceptible

components and identify revised primary and expansion inspection components, if needed

• Issue MRP-1911 update for SLR once screening inputs and inspection aging management strategies are complete

• Identify components where extended operation (80-100 years) could affect applicability of prior functionality analysis results (updates to MRP-2302/2323)

• Define and issue any interim guidance to lead PWR for any needed changes to Primary, Expansion and Existing components tables of MRP-227 Revision 14

• Brief NRC Staff and continue to engage with NRC in follow-up technical discussions in 2018-2019 time-frame

1 Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191) 10132342 Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP- 230, Revision 2) 10210263 Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232) 10165934 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1) 3002005349






Additional ActivitiesMultiple calls with NRC Research and the Division of License

Renewal including “deep-dive” Q&A callsMultiple meetings with NRC on GALL-SLR

– NRC has accepted many recommendations– Now keeping MRP-227-A1 as a reference for reactor internals AMP– Dropped the need for a BWR feedwater nozzle AMP– Expects new surveillance capsules to be inserted– All PWRs to perform volumetric examination on BMNs

Coordinated with LTO revising IAEA’s IGALL– 1st meeting in April 2016– No US participants involved on PWR IGALL work– Providing edits and updated references for AMPs and TLAAs

1 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) 1022863



Aging Management Roadmap


Enhancement and Modernization Technology Roadmap


For Reference - Summary of the SLR Approach & Communication PlanKey Elements (owner) Current Status

1 Continue the initial look at funding and potential budget flexibility (BWRVIP and MRP)

• EPRI identified $1M in additional funding for this effort through the "Strategic Gap Funding" process. MRP and BWRVIP eachallocated $500k to their proposed 2016 work plan.

• Estimate up to ~$500k/year additional from MRP could be applied towards PWR updates.

2 Discuss specific needs with Peach Bottom and Surry (BWRVIP and MRP)

• Meetings involving both planned lead plants have occurred.• BWRVIP representatives met with Peach Bottom personnel and made an initial assessment. After several follow-up discussions,

BWRVIP has what it needs for initial prioritization of their work. Where lead plant needs are not defined, the prioritization and updates will be developed based on generic fleet needs.

• MRP and NSSS vendors met with Surry representatives to assess needs and begin prioritization. The current position is that onlycore internals need to be addressed (i.e., only MRP-227 and various technical support documents). A plan was outlined for core internals. Proposals from Westinghouse and AREVA are now in-hand. The follow-up meeting with Surry occurred January 12, 2016.

3 Develop matrix of actual barriers to operation for the 26 identified products (BWRVIP and MRP)

• BWRVIP and MRP to work this in parallel, factoring in the learnings from item 2.• The current assessment from BWRVIP is that most inspection & evaluation documents related to reactor internals with apparent

time limitations include sufficient details regarding the underlying assumptions to the EOL conditions (e.g., neutron fluence, fatigue usage, hot operating time) to establish limitations associated with guidance applicability. It is anticipated that a standalone SLR aging management document could be produced to clarify these limits without the need to revise multiple guidelines on an accelerated schedule.

• MRP will attempt to follow a similar approach, but it is recognized that some of the data limits will be substantially more challenging and will require significant analysis (and time) to address.

4 Provide heads-up to NRC about our approach (Greenlee, Dyle) Discussions ongoing: RIC (Mar 8-10)and LRTF (Mar 11); at EPRI-NRC-RES coordination meeting (Taylor, Edsinger, Miller, Thomas). Additional meetings between LRTF and multiple meetings on GALL-SLR.

5Develop and maintain overall communication plan for use with NEI Licensing Renewal Task Force (Bernhoft), SLR Working Group (Taylor), and NSIAC (Edsinger)

• Tina Taylor to presented summary to SLR Working Group 11/6/2015.• Bill Pitesa provided update to NSIAC on 3/7/16.

6 Draft a work plan (with insights from item 2) and reconcile with current scope and budget (BWRVIP and MRP) • Complete plans were formally endorsed during NPC week (February 9-10, 2016). Handoff made to Materials Programs.

7 Draft a roadmap showing overall timing and coordination (primary system materials, but also concrete, cables, etc.) (Bernhoft) • Roadmaps have been updated for materials, cables and concrete, etc. and presented to MAPC 8/30/16

8 Support detailed discussion with NRC using insights from items 6 and 7 (Pitesa+Greenlee+Dyle+Bernhoft)

• Mike Gallagher (Exelon), Paul Aitkin (Dominion) , Jerud Hanson (NEI) and Sherry Bernhoft (EPRI) had a drop in visit with Chris Miller and others from DLR on 12/10/15. DLR was supportive of working with the industry to connect the 03-08 approach foraging management to SLR. Follow up mtg with NRC on 1/19/16 where industry further discussed use of MRP-227 in GALL-SLR

• SLR Executive Working Group including Bill Pitesa met with NRC 4/13/16. Routine interactions with NRC and LRTF as late as 7/28/16. Also plans to meet at technical levels as work for lead plants progresses.


Code and NRC


NRC Interactions, 1 of 3 02/17/19 – NRC/ACRS overview of GALL-SLR 02/19/16 – Initial discussion on MRP-227-A1 not referenced in GALL-SLR 03/02/16 – PWROG plan to transition to Master Curve 03/11/16 – License Renewal Task Force (LRTF) 04/12/16 – Annual NRC RES / EPRI management coordination meeting 04/13/16 – Quarterly tech call with materials leads 04/26/16 – GALL-SLR discussion

– NRC update on process of addressing 300 pages of public comments– Expressed willingness to find a solution for MRP-227 – NRC requested input on RPV issues– Unwilling to move on BMN inspections and BWR feedwater nozzle AMP

05/05/16 – PWROG met on CASS and internals cold work1 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) 1022863



NRC Interactions, 2 of 3 06/01/16 – GALL-SLR

– Industry offering options/challenges to GALL-SLR positions– NRC agreed to reconsider BMNs UT and BWR feedwater nozzle issues– NRC asked for additional clarification on several RPV issues

06/23/16 - Quarterly tech call with materials leads 06/26/16 – GALL-SLR

– NRC agreed to drop BWR feedwater nozzle AMP– NRC reiterated need for BMN baseline UT– NRC offered weak proposal on the need for additional RPV capsule

06/30/16 – Updated NRC on current environmental fatigue program elements and testing

07/19/16 – Baffle Former Bolt (BFB) meeting– BFB OE history– Safety Assessments– BFB focus group structure and plans


Recent Interactions, 3 of 3 07/28/16 – Materials executives and NRC management

– Industry executives, NRC directors and branch chiefs– Key topics included as part of program overviews

NDE Overview – Calvert Cliffs OE, DMW treated by MSIP GALL-SLR

– Good progress on MRP-227 and gap analysis– Concern over BMNs and additional RPV surveillance capsule

Need for NRC to resolve current RPV and CASS issues (BWR closed) Need for NRC to release FAVOR Code and positions on use of PFM NRC pleased with BFB response – wants frequent communication NRC will continue interaction on use of document screening tool Nuclear Promise and NRC budget (NRR restructuring)

08/04/16 – Annual NRC / Materials program tech exchange 08/17/16 – Steam Generator Task Force meeting 08/22/16 – NRC public meeting on Code limitations 08/23/16 – Drop-in on SLR issues - BMNs and RPV surveillance capsule


ASME Code Update Code meetings in May and August 2016 NRC participation in flux – numbers reduced and personnel uncertain Code Case N-853 approved - ½ nozzle repairs for branch connections Code case on excavate and weld repair (EWR) – One negative by NRC

– Changes to N-770-4 to account for these repairs under consideration. Revision to N-770 to match words in approved Version of N-729 received

NRC negative N-770-4 to provide inspection credit for peening of piping approved by ASME Other changes in progress for N-770

– Dissimilar metal weld branch connections (Pending PWROG analysis)– Revise cold led inspection requirements– OWOL examination frequency changes



© 2016 Electric Power Research Institute, Inc. All rights reserved.© 2015 Electric Power Research Institute, Inc. All rights reserved.

Drew OdellBWRVIP Chairman

Andy McGeheeSenior Program Manager, EPRI

New Orleans, LAAugust 31, 2016

BWRVIP Report -outMAPC Meeting



2016 BWRVIP Organization

Assessment CommitteeSteve Richter, Energy Northwest (Technical Chair)

Wynter McGruder, Xcel (Assessment Tech Vice Chair)

Denver Atwood, Southern Nuclear (Repair Tech Vice Chair)

Bob Carter, EPRI (Task Manager)

Mitigation CommitteeDan Miller, PPL (Technical Chairman)

David Willer, DTE (Technical Vice Chair)

Raj Pathania, EPRI (Task Manager)

Integration CommitteeDrew Odell, Exelon (Chairman)

Chuck Wirtz, EPRI (Task Manager)

Inspection Focus GroupChristian McKean, Exelon (Chairman)

Erica Mullen, DTE Energy (Vice Chairman)

Jeff Landrum, EPRI (Task Manager)

Executive CommitteeTim Hanley, Exelon (Chairman)

Mark Woodby, Entergy (Vice-Chairman)

Andy McGehee, EPRI (Program Manager) Executive Oversight Committee

Tim Hanley, Exelon (EC Chairman)

Mark Woodby, Entergy (AC Exec Sponsor)

Bill Kopchick, PSEG (MC Exec Sponsor)

Michael Kessler, KKL (International Sponsor)

Dennis Madison, Southern (BWROG Chair)

Dave Czufin, TVA (EOC at large)

Red = Vacant

Green = New

Mitigation Monitoring Focus GroupSteve Williams, Duke (Chairman)

Larry Loomis, Entergy (Vice Chairman)

Susan Garcia, EPRI (Task Manager)


BWRVIP Action Item from MACP Feb 2016 “BWRVIP to respond to the INPO challenge regarding dry tube

replacement (why it is a business decision and not a safety decision).” The BWR Safety Assessment – BWRVIP-06 documented failure of a in-

core housing and dry tube has no safety impact in the short term. The long term assessment stated, “Given the ability to detect significant

degradation and the ability to shut down safely with degradation, long-term actions are not required to assure safe operation.”

The BWRVIP acknowledges that dry tube cracking/failure does occur as described in GE SIL 409, revision 1.

Recommendations from INPO Materials Review Visits regarding the strategy for dry tube replacement is between the utility and INPO. The BWRVIP does not provide guidance for inspection, repair or replacement.


2016 BWRVIP Major Tasks

Address Jet Pump Flow-Induced Vibration Issues – Complete BWRVIP’s Full Scale Jet Pump Testing of BWR4’s Jet Pump Design

including supporting Vendor demonstrations

Support Hatch 1 and Dresden 3 with surveillance capsule activities BWRVIP SLR activities Extension of ISP for Subsequent License Renewal (SLR) Develop and review Materials IP Generic Topical Screening Process


2016 BWR Operating Experience

BWR 4 (Hatch 1) Spring Outage– Thin “Engineered” overlay showed an indication during repair activities.– Staff contacted by the utility– BWRVIP held Emergent Issue call and provided input to the utility– Full Structural Weld Overlay repair implemented as planned– BWRVIP surveyed the U.S. BWR fleet to see if any overlays of similar design

and materials existed and found there were none


1N2C and E Weld Original Overlay Design

6


PT after controlled grind

7


1N2C and E Weld Full Structural Overlay Design for 2016

8


2016 BWR Operating Experience

BWR 4 (Susquehanna 1) Summer Forced Outage due to drywell leakage

–Identified an ICMH leak near bottom flange–BWRVIP held Emergent Issue call and provided input to the utility–First of a kind leak location and repair–Utility performed a weld overlay on the affected area–The utility’s root cause analysis still in progress; upon completion, the BWRVIP

will review it for any generic implications


© Talen Energy 2014

View from Carousel Platform

10



Indication above Flange Weld

11



Flaw Characteristics – View of Indication

12


2016 Research Focus Areas RFA 1: Reactor Pressure Vessel Integrity SLR for BWR’s

RFA 9; Guideline Management Optimized VIP I&E Guidelines working through NRC staff approval

– VIP-18, Rev. 2 Core Spray – NRC SE received– VIP-41, Rev. 4 Jet Pump – NRC RAI response in progress– VIP-25, Rev. 1 Core Plate Bolts

Cast Austenitic Stainless Steel (CASS) work with the NRC staff – PWR and BWR effort– NRC SE received for BWRVIP-234 “BWR Vessel and Internals Project, Thermal Aging and Neutron

Embrittlement Evaluation of Cast Austenitic Stainless Steels for BWR Internals”

RFA 2: Thermal and Irradiation Effects on Stainless Steel Reactor Internals Crack growth and fracture toughness studies in high fluence BWR materials

‒ Zorita materials are currently being tested‒ Weld metal fracture toughness testing completed. Progress report to be issued in 2015‒ Crack growth rate testing still underway




Bernie RudellMRP Chairman, Exelon

Anne DemmaProgram Manager, EPRI


Materials Reliability Program



Presentation Topics

2016 Key Products Industry Challenges


2016 Key Deliverables

Topical Report for PWSCC Mitigation by Surface Stress Improvement (MRP-335 Rev. 3)

Basis for ASME Section XI Code Case N-838—Flaw Tolerance Evaluation of Cast Austenitic Stainless Steel (CASS) Piping Components (MRP-362 Rev. 1)

Effect of Lithium Concentration on IASCC Initiation in Irradiated Stainless Steel (MRP-413)

Management of Thermal Fatigue in Normally Stagnant Non-Isolable RCS Branch Lines (MRP-146, Rev. 2)

PWR Supplemental Surveillance Program (PSSP) Capsule Fabrication Report (MRP-412)


Industry Challenges Quasi-Laminar Flaws / Hydrogen Flaking Issue (Doel 3 RPV issue)

– MRP participated in the meeting organized by the Belgium regulator on this subject in January 2016

– MRP is assessing the impact of the new information on the previous probabilistic analysis (MRP-367) and will update report as needed

Potential Non-conservatism of Branch Technical Position 5-3 Methods to Estimate Initial Toughness of RPV Steels

– NRC agrees with MRP-401/ BWRVIP-287 that issue is not immediate safety concern but is performing analyses to determine if BTP must be revised

BMN Volumetric Examination– Agreement reached with NRC in 2013 to pursue Code Case to incentivize but not require volumetric

exams, however eventually tabled

Reactor Vessel Internals: Baffle-Former-Bolt Cracking– During spring 2016 outages, Indian Point 2 and Salem 1 found large quantities of degraded baffle-former-

bolts that required replacement, resulting in extended outages

– The findings raised concern regarding industry preparedness to respond when such degradation is discovered and the impact on the rest of the fleet


Industry Issue: Quasi-Laminar Flaws in RPV Forgings In 2012 UT ISI, quasi-laminar flaws were found at 2 Belgian PWRs (Doel 3, Tihange

2) in base metal areas of the forged beltline rings not usually inspected – Root cause: Hydrogen flaking, a fabrication defect– ~7,200 flaws in one forged ring at Doel 3

In 2014, reinspection of that ring using greater UT sensitivity identified ~13,000 flaws– Licensee demonstrated acceptability of flaws and both plants restarted in late 2015

In 2013 MRP assessed relevance of this Operating Experience by postulating the same [2012] condition in the limiting forged-ring PWR in the U.S. (MRP-367)– A probabilistic fracture mechanics (PFM) analysis was conducted for a vessel with 60 year

fluence and subjected to a severe pressurized thermal shock (PTS) transient– Through-wall cracking frequency = ~ 10-7 /year (met safety criterion of 10-6 /year)

NRC requested MRP-367 for information and cited MRP analyses as justification for not requiring special vessel inspections– NRC has requested ORNL to review MRP-367 (in progress); RAIs possible

MRP to update MRP-367 analyses for (1) higher number of flaws from 2014 inspection and (2) 80 years of operation


Industry Issue: NUREG-0800 Branch Technical Position 5-3 Fracture toughness reference temperature, RTNDT, is used to

evaluate RPV integrity per 10CFR50, Appendix G – Approximately 46 PWR vessels were fabricated before the ASME Code began

requiring, in 1973, the mechanical tests necessary to determine Initial RTNDT– In those cases, NUREG-0800 Branch Technical Position (BTP) 5-3 provides

acceptable alternatives for estimating RTNDT from other available test data In early 2014, some BTP 5-3 guidelines were identified as

potentially non-conservative MRP conducted joint project with BWRVIP to assess impact

– PFM analyses showed the fleet can continue to use existing BTP guidance though 60 years of operation with negligible increase in failure risk

– Based on review of MRP-401/BWRVIP-287 and a PWROG report (PWROG-15003-NP, Rev. 0), NRC agrees there is no immediate safety concern but is performing analyses to determine if BTP must be revised to address operation to 80 years

– NRC analyses to be completed by end of 2016


A600 BMNs susceptible to PWSCC; challenging to mitigate, repair, or replace; some significant inspection challenges exist

MRP safety assessments conclude routine visual exams are both necessary and sufficient for plant safety– Periodic volumetric exam is not necessary– Volumetric exam important for Extent of Condition determinations if a

leak is detected – Effective volumetric exam not presently possible for B&W unit BMNs

MRP ongoing BMN projects seek to– Ensure PWR fleet has the necessary tools available to assure safety and

address potential BMN degradation issues– Inspection methods and coverage requirements are available to satisfy

routine and issue-driven needs

Industry Issue: Bottom Mounted Nozzles (BMN) Volumetric Examination


Spring Baffle Former Bolts Issue: Industry Response

The Industry Baffle-Former Bolt Focus Group (BFB FG) was formed in May 2016 to support an integrated approach among industry organizations to address recent operating experience

- AREVA- EPRI- PWROG- Utility Staff- Westinghouse- Others

Six focus areas with key actions defined


Near Term Industry BFB FG Actions Completed

Supported presentation to NSIAC on 5/23/2016– Westinghouse Technical Bulletin TB-12-5 remains valid

Provided Industry Alert Letter from the PMMP Chairman to PWR site VPs on 6/1/2016– Expect that NEI 03-08 Interim Guidance will require the 4-loop plants

identified in the Westinghouse TB-12-5 bulletin to perform UT inspections of all the BFBs or replace an acceptable pattern of bolts at their next outage.

– Consideration should also be given to proceeding with procurement of replacement bolts prior to issuance of interim guidance due to potentially long manufacturing lead times.

Westinghouse NSAL 16-01 issued 07/05/16 AREVA CSB issued 07/14/16


Near Term Industry BFB FG Actions Completed

Issued NEI 03-08 “Needed” Interim Guidance on 7/25/2016 regarding BFB inspections for Tier 1 plants as identified in Westinghouse NSAL 16-01

Assessed Fall 2016 and Spring 2017 outage seasons for developing a contingency plan for tooling and BFB material needs– Fall 2016: 3 planned MRP-227 UT inspections (1 of 3 is a Tier 1a plant) and 1

VT-3 inspection (Tier 1b plant)– Spring 2017: 2 planned MRP-227 UT inspections (both Tier 1a plants), 1

planned UT inspection (non MRP-227 but a Tier 1a plant), and 1 VT-3 inspection (Tier 1b plant)

Initiated Hot Cell Post Irradiation Examinations of Indian Point 2 BFBs– Microscopic examinations have begun and are currently underway; may be done by 11/1/2016


Planned BFB FG Activities through Mid-2017

Finalize BFB OE database by adding international data and UT inspection results from 2016-2017 exams

Continue with Hot Cell PIE work for IP2 and SAL1Potential additional NEI 03-08 Interim Guidance for remainder

of U.S. PWR fleet (2-loop and 3-loop plants) later in Fall 2016 or early 2017

Establish fundamental understanding of BFB failure mechanism(s) and develop potential changes to MRP-227 inspection guidance as needed– Re-inspection frequency for UT exams– Allowance for proactive BFB replacement to manage aging




Jim StevensSGMP Chairman, Luminant

Helen CothronProgram Manager, EPRI


Steam Generator Management Program



2016/2017 Key Products

Generic Elements of U-Bend Tube Vibration Induced Fatigue Analysis for Westinghouse Model D5, Model 44F, and Model 51F Steam Generators

Triton Steam Generator Thermal-Hydraulics Code, Version 1.0 Steam Generator Examination Guidelines, Revision 8 Steam Generator Integrity Assessment Guidelines, Revision 4 Steam Generator In Situ Pressure Test Guidelines, Revision 5


2016+ Challenges

Assuring High Performance of Single Party Auto Analysis Systems

Foreign Object Wear– Detection of FOs

– Inspection Scope

– Estimating Wear rates from FOs

Cracking in Alloy 600TT Steam Generator TubesWear in Replacement Steam GeneratorsNeed for a State of the Art Thermal Hydraulic Code


Regulatory Issues

NRC has accepted SGMP Research Conclusions Regarding Divider Plate Crack Propagation– Staff Interim Guidance Letter

NRC will review published guidelines in 2016 to ensure all their technical issues have been resolved – Most of the technical issues have been incorporated into the guidelines;

however, not all of them are requirements

Single pass auto analysisSite-specific equivalency of generic eddy current technique

qualifications In plane fluid elastic instability


Operating Experience Update

South Texas identified a 0.09 volt volumetric indication in Alloy 690TT tubing during a steam generator inspection in 2015

Utility has completed the apparent cause and determined that if the indication is not a false call, the degradation is most likely pitting corrosion

They will be taking action in the future to address pitting corrosion in their hard sludge pile

SGMP will be conducting research 2017/2018 to investigate pitting corrosion in Alloy 690TT tubing



P R E S S U R I Z E D W A T E R R E A C T O R O W N E R S G R O U P

Global Expertise • One Voice

Bernie Rudell, MRP IC Chair (Exelon)August 31, 2016New Orleans, LA

PWROG Materials Committee SummaryMaterials Degradation and Aging Action Plan Committee

Meeting


Agenda Overview

• Significant Activities Since Last MAPC Meeting• Areas of Coordination & Strategic Planning with EPRI

MRP• 2016/2017 Products – Key Programs• 2016/2017 Challenges/Initiatives• Key Contacts

2


Significant Activities Since Last MAPC Meeting

3

Generic Projects

• Final SE received in May 2016 for WCAP-17096, “Reactor Internals Acceptance Criteria Methodology”. Working to issue the “A” version of the report.

• Issued final report PWROG-15032-NP, Revision 0, “Statistical Assessment of PWR RV Internals CASS Materials” (November 2015). The NRC made a formal request for the report for information only. Report was submitted in January 2016. Meeting held with the Staff on May 5, 2016. During the August 4th industry exchange meeting the NRC indicated that the RAIs discussed at the May 5th meeting were resolved and the Staff indicated they are finalizing their safety assessment.

• Issued final report PWROG-15105-NP, Revision 0, “PWR RV Internals Cold-Work Assessment” (April 2016). The NRC made a formal request for the report for information only and the report was provided in June 2016. Working to set up meeting with NRC.


Significant Activities Since Last MAPC Meeting –Cont’d

4

Generic Projects – Cont’d• Issued draft report PWROG-15089-P, Revision 0-A, “Plan for Transitioning RPV

Integrity to Direct Fracture Toughness” for review and comment in late January. The PWROG MSC and the program technical team met with the NRC in early March 2016 to present the plan.

• Issued final report PWROG-16025-NP, Revision 0 “Qualification/Refinement of Fluence Determination in Non-Traditional Reactor Vessel Beltline Locations for review and comment in March 2016.

• Issued draft report PWROG-15109-P, Revision 0-A “PWR Reactor Pressure Vessel Nozzle Appendix G Evaluation” for review and comment in July 2016.



5

Generic Projects – Cont’d• Issued draft report PWROG-16026-NP, Revision 0-A "Implementation of Phase 2

Recommendations for Three Representative Plants (PA-MSC-0551R2 Phase 3a)“ for review and comment in March 2016.



6

Cafeteria Projects• Issued additional draft and final reports under MSC-1283 “Evaluation of

Applicable Dissimilar Metal Welds Joining Alloy 600 Branch Connection Nozzle to Primary Loop Piping (B&W and Palisades only)” and MSC-1294 “Development of Contingency Weld Repair Design for Applicable Dissimilar Metal Welds Joining Alloy 600 Branch Connection Nozzles to Primary Loop Piping (B&W & Palisades Plants Only).

• Issued additional site specific reports under PA-MSC-0983 – “Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0.

• Issued final report PWROG-16003-P, Revision 0, “Evaluation of Potential Thermal Sleeve Flange Wear” (PA-MSC-1286).


Areas of Coordination & Strategic Planning with EPRI MRP

7

• Reactor Vessel Integrity Providing funding to defray costs associated with installation of PWR

Supplemental Surveillance Capsules at two plants

Working to demonstrate generically that nozzles are never bounding for P-T limits

Developed plan to transition industry and regulation to Direct Fracture Toughness (Master Curve)

o Meeting held on March 2, 2016 to present the overall plan


Areas of Coordination & Strategic Planning with EPRI MRP – Cont’d

8

• Reactor Internals Coordinating with the EPRI MRP on the Baffle-Former Bolt Inspection Findings

o Supported NSAL-16-1 and AREVA CSB 16-02 issuance and the 7/19 public meeting

Coordinating with EPRI MRP on MRP-227 LAI/RAI responses Supporting utilities in plant-specific applicability determinations, including MRP-191,

fluence, and cold worked stainless Working with MRP and NRC on a statistical approach for assessing CASS material in

PWR reactor internals

• Stainless Steel Degradation Working with the EPRI MRP on the development of I&E guidance for ID and OD

initiated SCC of PWR SS pressure boundary components


2016/2017 Products – Key Programs

• Programs to Support MRP-227-A - Reactor Internals PA-MSC-1403 - Scale and Impact of Uncertainty in Fluence Determinations for

Reactor Vessel Internals PA-MSC-1299 – Guide Card Wear RAI Support PA-MSC-1288 – PWR Materials Assessment PA-MSC-1103 - Functionality Analysis: Westinghouse Lower Support Columns PA-MSC-0983 - Support for Applicant Action Items 1, 2, and 7 from the Final Safety

Evaluation on MRP-227, Revision 0 (Working on plant specific requests) PA-MSC-0473 - Reactor Internals Acceptance Criteria Methodology & Data

Requirements (working to complete A-version of report)

9



10

• Programs to Support Reactor Vessel Integrity PA-MSC-1392 - Qualification/Refinement of Fluence Determination in Non-

Traditional Reactor Vessel Beltline Locations PA-MSC-1207 – Proactively Drive Changes in Reactor Vessel Embrittlement

Regulations PA-MSC-1200 - B&W Fabricated Reactor Vessel Materials and Surveillance Data

Information PA-MSC-1189 - Funding to Defray Costs Associated with Installation of Two PWR

Supplemental Surveillance Capsules PA-MSC-1182 - Revision to BAW-1543 for Master Integrated Reactor Vessel

Program PA-MSC-1091 – Demonstrate Excessive Appendix G Margins for PWR RPV

Nozzles PA-MSC-0783 - Archival of Westinghouse and CE Unirradiated Reactor Vessel

Materials



11

• Programs to Support Pressure Boundary Activities PA-MSC-1390 - Bottom Mounted Instrument Nozzle MNSA Repair Planning for

PWR Fleet – Task 1 only PA-MSC-1300 – PWROG Subsequent License Renewal PA-MSC-1294 - Development of Contingency Weld Repair Design for Applicable

Dissimilar Metal Welds Joining Alloy 600 Branch Connection Nozzles to Primary Loop Piping

PA-MSC-1283 - Evaluation of Applicable Dissimilar Metal Welds Joining Alloy 600 Branch Connection Nozzle to Primary Loop Piping (B&W and Palisades only)

PA-MSC-0551 - Development of Generic Recommendations to Address ID-Initiated and OD-Initiated SCC of PWR Stainless Steel Pressure Boundary Components (Phase 3a: Apply Phase 2 Recommendations to One B&W, One Westinghouse, and One CE Unit)


2016/2017 Challenges/Initiatives

• Resolving the NRC concerns with the MRP-227-A Guidance– Addressing and incorporating OE and Revision 1 impacts to existing OG products– Addressing A/LAI’s and recurring RAI’s

• Developing plan to transition industry and regulation to Direct Fracture Toughness (Master Curve)

• Working to address Baffle-to-Former Bolt inspection findings. Near term and longer term guidance being looked into.

• Supporting the Nuclear Promise

12


MSC PWROG Core/Planning Team Organization

13

Materials CommitteeHeather Malikowski, Exelon (Chair)

(610) 765-5864

Chris Wax, APS (Vice-Chair)(623) 393-6871

Reactor Internals Industry Planning TeamGlenn Gardner(860) 440-0373

Reactor Vessel Integrity Core TeamJim Nurrenbern, Ameren

(314) 225-1908

Pressure Boundary Core TeamSteve Petro, AEP

(269) 697-5048

Jim MolkenthinPWROG PMO(860) 731-6727

Tammy NatourAREVA

(434) 832-2763


Questions?

The Materials Committee is established to provide a forum

for the identification and resolution of materials issues

including their development, modification and

implementation to enhance the safe, efficient operation of

PWR plants.

14

Global Expertise • One Voicewww.pwrog.com


Kurt EdsingerDirector


Overall Training Strategy



Overview

Training being given increasingly higher priority at EPRI– Still needs to be training that connects to

core competencies of EPRI programs2016 training list available (3002006532)All available Materials training in the

program cockpitsContinuing to support Advanced

Engineering Training Initiative


Recent Progress

Nuclear Sector Training Matrix beta version available

PMMP piloting a knowledge transfer approach

MRP piloting video and CBT approaches for several high priority areas

BWRVIP piloting a program management platform approach that dovetails with training objectives

Beta Test Version


Overall Strategy Needed

Suggestion from David Czufin:Each program identifies their Top 20 topics for future

knowledge transfer– Arbitrary number (could be higher, but not much lower)

Members vote on the topics and prioritizePrograms determine the optimal mechanism for the

knowledge transfer (CBT, training, etc.)Programs develop, communicate and execute on a long-term

schedule

Time is right to step back and develop the detailed strategy



MAPC Meeting

David CzufinSr. VP Engineering and Operations SupportAugust 31, 2016

Knowledge Retention Program – Purpose & Process

Purpose: Means for ensuring continuity of business activities in light of potential workforce changes

Process Used:– Subject Matter Expert Documented Knowledge/Skill Loss on an Electronic

Interview Form– Compiled a list of Knowledge/Skill Loss Items using a small team– Developed a preliminary Knowledge Retention Action Plan

Next Steps: Implement, Monitor, and Evaluate Action Plans and Priorities

– Identify action owners, target completion and issues identified

2

Knowledge Retention & Transfer

Methods for Knowledge Retention and Transfer

StaffingProcess Re-Engineering

Alternative or Shared ResourcesEducation and Coaching

Documentation and Codification

3

Unique Knowledge, Experience & SkillsTypically uncodified information gained through the interaction with co-

workers, mentors, supervisors, etc., usually hard to catalog or document in detail. Often based on judgment and informed action.

Working Level Knowledge, Experience & SkillsKnowledge typically obtained through books, documents, formal training, and

experience applying the knowledge.

Foundational KnowledgeFamiliarity usually gained through education, experience or association

focused on a particular body of study (e.g., engineering design principles, etc.)

Types of Knowledge

Robin Dyle - Knowledge and Skills Summary

4

Unique Knowledge, Experience & SkillsASME Code Application & Committees

Working Level Knowledge, Experience & SkillsTechnical Underpinnings of BWRVIP and MRP; ASME; Code of Federal

Regulations; The GALL; Corrosion & Chemistry; NEI 03-08; Working Knowledge of Fracture Mechanics, Risk, Fatigue, Metallurgy and

Welding.

Foundational KnowledgeEngineering/Technical Degree; Plant Systems (Tech Specs, Plant Ops);

EPRI IMT & MDM; History of BWRVIP and MRP; Leadership & Effective Communications – Big Picture/Strategic Perspective, Ability to

effectively interface from Executives to Engineers

“Education is not the learning of facts, but the training of the mind to think.” – Albert Einstein

Skills Retention Action Plan - Foundational

Foundational KnowledgeEngineering/Technical Degree; Plant Systems (Tech Specs, Plant Ops); EPRI IMT & MDM; History of BWRVIP and MRP;

Leadership & Effective Communications – Big Picture/Strategic Perspective, Ability to effectively interface

from Executives to Engineers

5

Knowledge/Skill Necessary Skills or Actions to Acquire• Engineering or Materials Science Degree • General knowledge of plant Reactor Theory and Systems• Technical Specifications – Application & knowledge related to licensing and

operation, and Reactor systems • EPRI IMT and MDM – general knowledge of use and applications• BWRVIP – history and program owner level knowledge • MRP - history and program owner level knowledge • Leadership/Effective Communications proficiency

Proposed Solutions:Utility Plant Systems TrainingUtility Technical Specifications Overview Training (more specific to Rx)EPRI IMT and MDM Training (from past training & presentations)BWRVIP history and program manager training (from previous training, presentations, Cockpit; consider Chuck Wirtz or Ken Wolf as resource)MRP history and program manager training (from previous training, presentations, Cockpit; consider Tim Wells, Jack Spanner, Craig Harrington)

Skills Retention Action Plan – Working Level

Working Level Knowledge, Experience & SkillsTechnical Underpinnings of BWRVIP and MRP;

ASME; Code of Federal Regulations; The GALL; Corrosion & Chemistry; NEI 03-08; Working

Knowledge of Fracture Mechanics, Risk, Fatigue, Metallurgy, NDE and Welding.

6

Knowledge/Skill Necessary Skills or Actions to Acquire• BWRVIP – assumptions, bases & knowledge of key BWRVIP documents• MRP – assumptions, bases & knowledge of key MRP documents• ASME – assumptions, bases & knowledge of key Codes & Codes Cases• Code of Federal Regulations – knowledge/understanding of key CFR’s• NEI 03-08 – historical perspective and current process• GALL - general understanding of license renewal regulatory aspects and

unique technical requirements needed to satisfy beyond base programs. • Corrosion – drivers such as environmental effects on material, chemistry,

and how to use corrosion research to close gaps.• Chemistry – mitigation, primary water chemistry and why it is controlled• Knowledge/Experience - Fracture Mechanics, Risk, Fatigue, Metallurgy,

NDE, and Welding

Proposed Solutions:Utilize EPRI, Industry and Vendor SME’s to develop & provide training and/or training materials to close gaps

Skills Retention Action Plan - Unique

Unique Knowledge, Experience & SkillsASME Code Application & Committees

7

Knowledge/Skill Necessary Skills or Actions to AcquireASME Code:• Understand how code is applied at the plant vs. understand how to process

something using the code – weigh the advantage of each• Need to affect the policy in the direction that the code goes• Most advantageous way to apply the code

Proposed Solutions:ASME Section III and XI trainingLeverage Nathan Palm’s background, working knowledge of code committees and policy direction


Anne DemmaProgram Manager, EPRI


Progress on Knowledge Transfer / Training

Examples from MRP



General Approach

MRP identified training topics of interest, prioritized, and decided to develop 3 training modules in 2016-17

MRP will identify other top topics for future knowledge transfer

MRP members will vote on the topics and prioritize themFor each of the topics, we will determine the recommended

mechanism for the knowledge transferWe will develop a long-term schedule with topics and when

the knowledge transfer will be completed


Current Examples from MRP

Developing 3 computer-based training modules on– Thermal Fatigue – Reactor Internals Management– Reactor Vessel Integrity

Awarded contract to experienced CBT developer– NANTEL requirements – Numerous CBTs based on EPRI products


MRP Approach to Developing Training Content for CBTs identified

– Video taped workshops held at EPRI LWR Materials Reliability Conference in Chicago on August 1

– Transmitted all EPRI reference materials and products that are needed for CBTs to developer

Scope, deliverables and schedule defined– Storyboards– Scripts and study guides– Draft CBT modules– Final draft CBT modules for testing– Final CBT modules for publication– NANTEL CBTs and Test Banks

CBT developer attended workshops Two EPRI Materials staff trained on e-learning and training software used by

CBT developer (i.e., Storyline 2)




Andy McGeheeSenior Program Manager


Technology Transfer Concept

BWRVIP “Turbo Tax”



Current Way We Transfer Technology (Guidance) Assessment Inspection Repair/Replace Mitigation

Component (I&E) Guidelines Guidelines Design Criteria RecommendationsCore shroud BWRVIP-76, R1-A BWRVIP-03 BWRVIP-02-A/-04-A BWRVIP-62, R1/-190, R1

Core spray BWRVIP-18, R2 BWRVIP-03 BWRVIP-16-A/-19-A/-34-A N/A

Shroud support BWRVIP-38 BWRVIP-03 BWRVIP-52-A BWRVIP-62, R1/-190, R1

Top Guide BWRVIP-26-A BWRVIP-03 BWRVIP-50-A N/A

Core Plate BWRVIP-25 BWRVIP-03 BWRVIP-50-A BWRVIP-62, R1/-190, R1

SLC BWRVIP-27-A BWRVIP-03 BWRVIP-53-A BWRVIP-62, R1/-190, R1

Jet pump assembly BWRVIP-41, R3 BWRVIP-03 BWRVIP-51-A BWRVIP-62, R1/-190, R1

CRD guide/stub tube BWRVIP-47-A BWRVIP-03 BWRVIP-17/-55-A/-58-A BWRVIP-62, R1/-190, R1

In-core housing/dry tube BWRVIP-47-A BWRVIP-03 BWRVIP-17/-55-A BWRVIP-62, R1/-190, R1

Instrument penetrations BWRVIP-49-A BWRVIP-03 BWRVIP-57-A BWRVIP-62, R1/-190, R1

LPCI coupling BWRVIP-42, R1 BWRVIP-03 BWRVIP-56-A N/A

Vessel ID brackets BWRVIP-48-A BWRVIP-03 BWRVIP-52-A BWRVIP-62, R1/-190, R1

Reactor pressure vessel BWRVIP-74-A N/A N/A N/A

Primary system piping BWVIP-75-A N/A N/A BWRVIP-62, R1/-190, R1

Steam dryer BWRVIP-139-A BWRVIP-03 BWRVIP-181-A N/A

Access hole cover BWRVIP-180 BWRVIP-03 N/A BWRVIP-62, R1/-190, R1

Top guide grid beam BWRVIP-183 BWRVIP-03 BWRVIP-50-A N/A

Bottom head drain line BWRVIP-205 N/A BWRVIP-208 N/A


Preferred Way to Transfer Technology


Concept

The vision is to package the documents from the BWRVIP program into an electronic format that:– Connects all of the program pieces (Inspection, Evaluation, Mitigation, Repair/Replacement,

Tech Bases)– Tailors the guidance to a utilities specific plant– Simplifies the task of building and managing a program– Reinforces the connection between the information and the plant (reinforces the connections

related to what and why)– Provides a user friendly interface for all program interactions– Meets all BWRVIP program requirements– Is always up to date– Helps EPRI staff understand implementation, which will ultimately allows us to better

understand gaps and develop more efficient solutions


What do we want the tool to do

Build a Program Track & inform on required inspections (provide all relevant references used in algorithms) Track inspection intervals / frequencies per current guidance Provide / Suggest Inspection options per current guidance (maximize efficiencies and

flexibility for integrated outage management plans – critical path, etc) Store detailed inspection results (weld ID, locations, parameters, pictures, video, etc) Generate Inspection Report to report to VIP (for inspection trends report) Provide applicable flaw evaluation guidance and store completed evaluations and any flaw

handbook info Alert user if deviations (like turbo tax audit flag) may be necessary based upon inputs Track INPO Review visit outcomes (AFIs /strengths, etc) with applicable components /

activities X…? Y…? Z…?


Thoughts – Comments - Questions



© Copyright 2016 Institute of Nuclear Power Operations

NPC MAPCINPO UpdateCarl Larsen

Nuclear Asset Protection


Outline• Recent important events

• Delivering the Nuclear Promise and engineering programs


Recent Important Events - PWRs1. Baffle former bolt degradation, emergent replacement, outage extensions at two plants

2. First potential corrosion degradation mechanism in 690 TT SGs

3. Pressurizer safety relief valve weld flaw, 80% through-wall

4. Unit shutdown due to RCS leakage from CVCS drain line due to lapses in BACC program

5. CRGT guide card replacements due to wear at international plants (201 of 1178 CRGTs)

6. Missed flaw detection with array probe in tubesheet (two 600 TT plants)

7. RHR suction line branch connection to hot leg 8” long weld indication – scope expansion

8. Through-wall RCS (CVCS) leak

9. Through-wall RCS (HPSI pump suction casing vent valve) leak


Recent Important Events - BWRs1. At least two more shrouds identified with off-axis

(atypical) cracking

2. Under-vessel LPRM leakage

3. Recirc system overlays


DNP and Engineering Programs• Transfer of many responsibilities to

corporate

– Benefits

• Consistency

• Use of best SMEs

– Challenges

• Station influence (org chart dotted lines, budgets)

• Implementation monitoring and oversight


DNP and Engineering Programs• August 2016 Engineering Programs

Managers Workshop at INPO

– Breakout session: What program owner behaviors/roles/responsibilities will need to change or be re-enforced as a result of the Nuclear Promise?


Actions/Options Going Forward• Engineers have to understand fiscal

responsibility– Know costs and resources for each job

– Understand and communicate risk better

– Understand that best option is not always the most conservative option

• Need early identification of industry issues

• Owners need to stay in their assigned roles and responsibilities


Barriers• Fully understanding the components of risk

– Applies to nuclear safety, generation, asset protection

– Probability of occurrence and consequence (including enterprise risk)

• Dependence on Engineering to work outside of core business


Current Best Practices• Communicate that senior leaders share risk

• Obtain site, stakeholder buy-in for assumed risk

• Roles/responsibility matrix

• Advocacy sharing between program and system engineering

• Good industry group engagement

• Be efficient without sacrificing technical conscience


Programs and Technical Conscience• How is Program Engineering different?

– Events (lagging indicators) appear to be few and far between

– Trending is difficult

– Events tend to be consequential

– Focus needs to be on barriers to events

– It might be years before weakened barriers result in an event


2010:DC Cook 2

baffle formerbolts

2006:Cruas 4

SG tube burst

Emergent Repairs

2007:ANO 1

tie-rod bowingin new SGs

2005:Oconee 1, 2, 3tube wear in

new SGs

2009:SONGS 2

piping ODSCC

2008:Laguna Verde 1

jet pumpdamage

2010:Davis-Besse

RPV head again

2005-2009:Limerick 2

jet pumprestrainers

2006:Hatch 1

shroud tie-rod

2010-’15:Susquehanna 1

jet pump restrainers

2010:Sizewell B

pressurizer heater sleeve

2010:Bugey 3

emergent SGreplacement 2016:

IP 2, Salem 1baffle former

bolts

2013:Palo Verde 3

BMI leak


2000:VC Summer

DM weld leak

1992:Brunswick 1

shroud IGSCC

Extended Shutdowns

1996:Salem 1

emergent SGreplacement

1984:Nine Mile 1

recirc pipingIGSCC

1998 -2010:12 Japanese

shroudreplacements

2003:STP bottom headpenetration leak

2000:Indian Pt. 2replace SGs

1982:Trojan

baffle jetting,fuel damage

2002:Davis-BesseRPV head

1986:Pilgrim

recirc pipingIGSCC

2012, 2014:Doel 3, Tihange 2

RPV fracturetoughness


2002, 2003:Quad Cities 2

steam dryer failures

1997:Peach Bottom 3

jet pumpriser crack

Forced and Midcycle Outages

2002:Quad Cities 1

jet pump beam

1993:Palo Verde 2

SG tube burst

2003:Quad Cities 1

steam dryer failure

1993:Grand Gulf

jet pumpmixer ejection

2013:Oconee 1

HPI leak, thermal fatigue

2014:Robinson SGpri/sec leak

2012:San Onofre 3

SG pri/sec tube leak

2001:Palisades

CRD unisolableleak

2014:Harris CRDM

UT data re-analysis

2007:Waterford 3SG batwings


Unit Ending Events

1991:Yankee Rowe

RPV

1997:Maine Yankee

SGs

1992: Trojan

SGs


• Good to look for efficiencies in engineering programs, but use Technical Conscience

• Engineering programs events can be consequential• Would some of our events a few years ago be unit-ending

events today?• INPO will still evaluate for safety and reliability, per PO&Cs

Summary


Jeff Goldstein, Program Chair, EntergyDan Wells, Program Manager, Chemistry

Materials Action Plan Committee MeetingWednesday, August 31, 2016

Water Chemistry ProgramMAPC Update



Chemistry Research Focus Areas

Chemistry Guidance (Guidelines, Sourcebooks)

Management of Corrosion Product Behavior and Impacts

Chemical MitigationRadioactivity Generation and Control (Source Term Reduction)

Chemistry Monitoring and Control

Joint with RSChemistry Modeling(Fundamental)

Chemistry Benchmarking and Trending (Fundamental)

Fundamental RFAs – Not Prioritized

RS = Radiation Safety Program


2017-2018 Water Chemistry PortfolioIncluding TSG and Non-Chemistry Program Funded Work

Chemistry Guidance (Guidelines,

Sourcebooks)

PWR Secondary Chemistry Guidelines Revision 8

(2015-2017)

Revision to the Condensate Polishing Guidelines

(2016-2018)

BWR Water Chemistry Guidelines Rev. 8

(2018-2019)*

Open Cooling Water Guidelines Review (2017)*

Risk Informed Chemistry Control (2017-2018)*

Chemistry Control for Flexible Power Operation

(2015-2018)

Chemical Mitigation

Effect of Amine Decomposition Products on

Crack Growth Rates (2017-2019)*

Hydrazine Alternatives: Demo (2018-2019)*

Qualification of KOH for Plant Trial (2017)*

Li-7 Recovery Technology (2015-2017)

Hydrazine Alternatives: Current Tech Assessment

(2016-2017)

Management of Corrosion Product

Deposition and Transport

PWR Secondary Side Filming Amine (FA)

Application (2016-2017)

Dispersants: SG Deposit Evaluation (2017-2018)*

Filming Amine Qualification Testing (2018-2019)*

Impact of Fuel Materials Changes

(2018-2019)*

Radioactivity Generation and

Control (Source Term Reduction)

Micro-Environment Effects(2015-2017)

Surface Passivation of Primary Components

(2015-2018)

Hydrophobic Coatings for Contamination Control in

NPP (2016-2017)

Behavior of Ag and Sb(2016-2018)

Optimization of Zinc for Benefits and Cost

(2018-2019)*

Davis-Besse Gamma Scan Following Zinc (-2017)

Chemistry Monitoring and Control

On-Line Monitoring of Anions

(2016-2017)

On-Line Iron Analysis: Demo (2018)*

On-Line Iron Analysis: Tech Assessment (2016-2017)

Modeling of Multiple Alkali Chemistry (KOH)

(2016-2017)

Evaluation of Optimized Sample Frequency

(2015-2017)

Silica Quantification in BWRs: Demo (2015-2017)

Base Funded Work New* TSG (Supplemental) Funding Other Program Funding

Projects with Materials Related Scope


2017+ Chemistry Guidelines Review and Revision

Water Chemistry

Control Guidance

BWR Water Chemistry Guidelines Revision (2018-2019)• Incorporate SIGNIFICANT

interim guidance for chloride AL1 and mitigation monitoring

• Address application to advanced designs and flexible power operation

PWR Secondary Guidelines publication (2017)• Final draft with committee• Comments requested by August

29

Open Water Chemistry Guidelines (2017)• First version published in 2012• Review to identify if revision is

necessary• Understand industry application

Risk Informed Chemistry Control (2017-2018)• What if operation for an

additional 20 years isn’t an objective?

• Could cost be reduced with alternative chemistry control under prerogative of economic hardship?


BWR Water Chemistry Interim Guidance Applicable to BWRVIP-190, Rev. 1

1. IGSCC mitigation monitoring at BWRs operating with noble metal hydrogen water chemistry, in light of issues identified with the measurement of electrochemical corrosion potential (ECP) at the external Mitigation Monitoring System (MMS)

– Six (6) Needed elements, four (4) Good Practices

2. Revise Action Level 1 for Reactor Water Chloride

– Response to laboratory testing showing accelerated SCC crack growth rates in 288°C oxygenated BWR water with chloride as low as 3-5 ppb One (1) revised Needed element – AL1,

power operation >10% to ≥3 ppb – all chemistry regimes

Topic StatusReview period for MMFG (document and response form sent by email on Nov. 9th) Complete

Finetech/EPRI address comments/incorporate revisions Complete

Three week review period for Mitigation Committee Complete

Finetech/EPRI to address comments/incorporate revisions (includes BWRVIP MMFG webcast) Complete

Two week review period for Integration Committee Aug 8 - 22

EPRI/Finetech address comments/incorporate revisions Two weeks

Two week review period for Executive Committee Two weeks

EPRI/Finetech address comments/incorporate revisions Two weeks

Issue official Interim Guidance letter with implementation required July 2017 (approx. 9 months from issuing)

Target 4Q2016

publication

Plant Implementation After spring 2017 RFOs


Steam Chemistry Project Update

EPRI cooperating with steam turbine vendors to develop appropriate technical basis and guidance

Challenge: achieving both needed chemistry for the balance of plant and the SGs, and the needed chemistry for the steam turbines. Specifically use of amines to reduce FAC and corrosion product transport to SGs and maintaining required steam chemistry for turbines

• Gap: Are amine degradation products (e.g., acetate and formate) detrimental to turbine materials?• Turbine warranties may be affected

Co-funded by Chemistry, SGMP, and EPRI Generation

• Testing at 3 temperatures and 4 chemistries

• Seven materials (rotor & blade materials)

Phase 1: Pitting

Testing –underway

(2015-2016)

• Test Plan by Phase 1

• Evaluation ~8 materials, 3 environments

Phase 2: Crack Growth Rate Testing (2017-2019)

• Cation conductivity

• Quantification of organic acids

Phase 3: If Necessary -Update PWR Secondary Guidance


Corrosion Product Mitigation Technologies

Fe2+

Filming Amine (FA) Technology

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

N

H HCH

Full Secondary System

H2O

Base MetalCS, SS, Ni-based alloys

coolant Fe2+

loosely adherent Fe deposits

Dispersant (PAA) Technology

SGBalance of Plant

FexOyFexOy


Status of Corrosion Product Mitigation TechnologiesCurrent and Planned Work

Dispersant Technology

Three (3/4) different applications– Startup, Online, SG wet layup, Hybrid

application at end of cycle / shutdown

2016 workshop planned for 22-23 September 2016 (conjunction with SG Sec. Side Management Conf., Orlando, FL)

New Project (2017-2018) to understand plant specific thermal performance changes

Filming Amines – New Technology

Filming amines have been applied at fossil plants for over 25 years Protect carbon and low alloy steel components,

especially during periods of long layup Recently applied at the Almaraz Nuclear Power

Plant in Spain and tested at Embalse in Argentina

EPRI Project Scope– Phase 1: Technology Evaluation (2015-2017)– Phase 2: Qualification Testing (2018-2019)– Phase 3: Plant Demonstration (2019+)

Active collaborations with AREVA GmbH and COG




Jim Cirilli (Exelon), PSCR TAC Chair

TG Lian / Anne Demma (EPRI), PSCR Program Manager

Materials Degradation & Aging Action Plan Committee MeetingAugust 31, 2016; New Orleans, LA

Program Summary Primary System Corrosion Research



Alignment of Research FocusProject Schedule Issue Programs RFAs

PSCR Research Focus 2016-17 Project(s) & 2018 Proposals 2016 2017 2018 2019+ IMT gaps MRP BWRVIP SGMP WRTC

RFA-01: IASCC of InternalsDevelop quantitative understanding in IASCC occurance in BWR, PWR and VVER reactors to address IASCC concerns

CT size and orientation effects on IASCC

P-AS-14aP-AS-14bB-AS-10B-AS-26B-AS-11

RFA-2

Effects of high fluence neutron irradiation on localized deformation and IASCC RFA-1 RFA-2

Develop technical basis for solutions to counter IASCC RFA-1 RFA-2

Support Core Shroud cracking investigation RFA-2

Development of IASCC models for Type 321 Ti-stainless steels in VVER RFA-1

Investigate effects of heat-to-heat variability on IASCC susceptibility and the effectiveness of HWC to mitigate IASCC

RFA-1 RFA-2

Microstructural Contribution to IASCC Susceptibility of Baffle Bolt Stainless Steels RFA-1 RFA-2

RFA-02: Void Swelling of InternalsDevelop high-fidelity void swelling assessment capabilities for core internals in PWR and VVER

Rapid Simulation of High Fluence -- Ion radiation of LWR irradiated FTT P-AS-15 RFA-1

RFA-03: Irradation Embrittlement of SS, SS welds, and CASS in reactor internalsContinue to improve the understanding of the effects of increasing fluence on the embrittlement of reactor internal components

Modeling of irradiated mechanical properties & Toughness P-AS-13aP-AS-13bB-AS-09B-AS-12 P-AS-38

RFA-1RFA-2

Small-volume mechanical property evaluation RFA-2

Round Robin – APT Data Acquisition and Analysis RFA-2

Correlation between irradiation microsturcture and mechanical properties for SS and SS weld RFA-1

RFA-04: IASCC and fracture toughness of irradiated Weld repairs

No project To coordinate with WRTC

B-RR-02P-RR-03

BWRRFA-5

WRTCRFA-2

RFA-05: Embrittlement of RPVsImprove understanding of potential RPV embrittlement mechnisms

No project To coordinate with PWR and BWR RPV integrity focus

program

P-AS-47P-AS-04, 05

P-DM-10B-AS-05 , 36

RFA-9 RFA-1


Project Schedule Issue Programs RFAs PSCR Research Focus 2016-17 Project(s) & 2018 Proposals 2016 2017 2018 2019+ IMT gaps MRP BWRVIP SGMP WRTC

RFA-06: Next Generation MaterialsDevelop materials with better resistance to irradiation induced degradation such as IASCC, void swelling, and embrittlement, etc.

Advanced Radiation Resistant Material (ARRM) Phase-I P-RR-08 MRFA-1

MRFA-2ARRM Phase-II (2018 - 2023)

RFA-07: PWSCC in Ni-based AlloysImprove mechanistic understanding in PWSCC of Ni-based alloy and enhance confidence in long term reliability of Alloy 690/52/152 components in PWR.

Determine distinct PWSCC dependence (effects of microstructures in Alloy 600) P-AS-11

P-AS-12P-AS-20P-DM-13

RFA-3

Investigate PWSCC of Alloy 600 through GB oxidation and GB cohesive strength RFA-3

PWSCC of Alloy 690-52-Carbon Steel Welds in CANDU primary heat transfer system

RFA-3 (Candu)

RFA-08: ODSCC and PbSCC of SG TubesAssess the overall significance of ODSCC and PbSCC on steam generator tubes (Alloy 600TT, 800, & 690TT)

SGMP-PSCR Joint Proposal -- Investigate SCC mechanisms of Alloy 600TT SG tube via microstructure characterization

P-AS-24P-AS-30P-AS-41P-AS-42P-DM-13

Corrosion mech

RFA-09: SCC in LASQuantitative determination of increase in SCC crack propogation in low alloy steels.

Effects of irradiation hardening on SCC CGR in LAS microsturcture

B-DM-07 B-DM-08 RFA-1

RFA-10: Environmental FatigueDevelop mechanistic understanding of EAF

Mechanistic understanding of loading effects on EAF CGR (end in 2016)

P-AS-02B-AS-07B-AS-14 P-AS-16

RFA-10 RFA-10

Investigation the effects of irradiation on EAF CGR (end in 2016) RFA-10 RFA-10

Short crack behavior in EAF of stainless steel RFA-10 RFA-10

RFA-11: Environmental FractureInvestigate the environmental effects on fracture resistance

No project To coordinate with PWR and BWRVIP

P-DM-09P-AS-40B-DM-06 B-DM-03

RFA-9 RFA-1

RFA-12: Thermal AgingPerform thermal aging studies for CASS, LAS, martensitic stainless steels, and other materials

No projectP-AS-46P-DM-15P-DM-16

RFA-4

Alignment of Research Focus


Key 2016/2017 Products

Investigation of the irradiation effects on EAF of stainless steels (2016)• Will provide results to BWRVIP and MRP to improve aging management strategies for reactor

internal components

The effects of loading and composition on EAF crack growth (2016)• Will lead to additional information for ASME code N-809

Investigate PWSCC initiation of Ni-based Alloy (2016/2017/2018) • The results of this overall project can be used as:

a) Technical basis to demonstrate the excellent PWSCC resistance in Alloy 690b) Technical basis to improve current Alloy 600 management strategies

Development of IASCC crack growth model for Ti-stainless steels (2017 -- )• Improve aging management for VVER reactor core internals• Utilize EPRI research results on IASCC for Western PWR


Study Irradiation Effects on EAF Crack Growth Rates

Main Issue:• RFA-10: Environmental Fatigue, IMT gaps: P-AS-02 &16; B-AS-07 & 14• Address a knowledge gap whether irradiation can enhance EAF crack growth rates• NRC has requested EPRI to provide more information as part of overall EAF issue

Research Approach:• Use the IASCC database and other data sources to capture all CGR information associated cyclic

loading• No additional testing is involved in this effort• Develop a theoretical model on EAF of irradiated stainless steel

How to Use the Result:• Will transfer the theoretical model to BWRVIP and MRP for follow on work and implementation • If necessary will assist issue program to pursue ASME code case (as we did for IASCC crack growth

curves for irradiated stainless steels) Project Status:

• A final report and theoretical model will be published at the end of 2016


Summary: Irradiation Effects on EAF CGR

Longer rise-time ( tr > 50 s) cyclic data from IASCC tests appear to be usable for modeling EAF crack growth in irradiated materials (some shorter rise-time data are also usable)

The enhancement over air fatigue appears to be similar to non-irradiated material plus an additional effect of dose or irradiated yield stress

Identified and collected EAF data that can be used for modeling CGR of irradiated core internals materials

Develop models for predicting EAF crack growth rate of irradiated austenitic stainless steel materials (to be completed in 2016)


The Effects of Loading and Composition on EAF Growth Main Issue:

• Understand EAF growth behavior under loading transients (rise and hold times) in PWR• Understand the effects of composition (S content) on EAF growth rate• Investigate reducing conservatism in pending ASME code N-809 (Bettis model) on EAF growth rate• Generate information to be used in total fatigue life assessment

Research Approach:• Consider a wide range of rise and hold times• Cover a wide range of S content in stainless steels used in PWRs• Leverage the experience of AMEC / Rolls Royce from their previous and on-going EAF research in UK

How to Use the Result:• Will provide the information to EPRI-coordinated international EAF working group (EDF, Rolls Royce,

AREVA, EPRI)• Collaborate with BMPC (Bettis) to pursue ASME code case with reduced conservatism

Project Status:• A final report will be published at the end of 2016


The Effects of Loading and Composition on EAF Growth

The degree of enhancement of EAF growth rates in PWR water vs. air depends on loading transients and sulfur (S) content of stainless steels

Higher S steels show less enhancement of EAF growth rates vs air than lower S steels

The retardation effects with S are greater for longer rise and hold times

These results offer the possibility of reducing conservatism in existing EAF growth rate models such as code case N-809


Investigate PWSCC Initiation of Ni-based Alloy Main Issue:

• The technical basis to take credit for the excellent PWSCC resistance of Alloy 690• The lack of ability in predicting PWSCC occurrence in Alloy 600 components

Research Approach:• Perform parametric PWSCC studies (EPRI-NRC-PNNL, EPRI-GEGRC, etc)• Perform grainboundary characterization (for GB oxidation, cohesive strength, and threshold stress)• Co-fund with NRC and EDF

How to Use the Result:• Will provide the results and participate in the development of EDF PWSCC Local model for plant

aging management • Provide the mechanistic basis for factor of improvement of Alloy 690 over Alloy 600

Project Status:• Testing started in 2015, GB analysis starts in 2016• Complete the project in 2018


Ti-Type 321 SS IASCC model

Main Issue:• Provide technical basis in order to develop IASCC crack growth models model for VVER internals

o current IASCC models in VERLIFE for VVER are based on Russian knowledge, which has not been sufficiently open or communicated with VVER stake holders

Research Approach:• Expand existing EPRI IASCC models (on 304SS and 316SS) to Ti – 321SS

o Compile and compare CGRs of Irradiated type 321SS with EPRI database/modelo Identify data gapso Perform IASCC CGR tests to fill the gaps (if necessary)o Develop IASCC model of Ti-321SS for VVER

How to Use the Result:• Hand off to VVER members to improve or replace CGR models in VERLIFE

Project Status:• Data and gap analysis in 2017; If needed, testing in 2018-19• Complete project in 2020 (or sooner without testing)


EPRI IASCC CGR Model (Report 3002003103)

Model vs. measured plot for PWR model and calibration data at 310ºC,showing a reasonable fit overall


2016 Deliverables1. 2016 Revision of Materials Degradation Matrix (MDM) (Due: 12/20/16; 3002007968)

2. Environmentally-Assisted Fatigue Crack Growth in Irradiated Stainless Steels in LWR Environments (Due: 12/20/16; 3002007969)

3. Summary of Investigation of the effects of loading on Environmentally-Assisted Fatigue Crack Growth in Stainless Steels in LWR Environments (Due: 12/20/16; 3002007973)

4. Development of Small-volume Mechanical Testing Methodology (Due: 8/31/16; 3002007970)

5. Final Report of POLIM Project (Due: 8/31/16; 3002007971)

6. (EPRI-LWRS project) Establishment of the Capability to Determine IASCC Initiation from Four-Point Bend Tests Conducted in BWR NWC Environment (Due: 10/31/16; 3002007972 )

7. IASCC Susceptibility and Evolution of Microstructure in Several Ni-Base Alloys after Proton Irradiation (Published; 3002007461)


Key 2016/2017 Challenges

Develop a short-term investigative strategy to understand the underlying cause of IASCC occurrences in BWR and PWR core internals• To achieve mechanistic understanding of the nature of cracking in core shroud (BWR) and in

baffle-to-former bolts (PWR)o To complement the current failure analysis efforts

• To provide in-depth insights to o BWR fleet on performance of core shroud and top guide o PWR fleet on reliability of high strength bolts used in core assembly

Develop a cost & time effective plan to understand the IASCC resistance in various types of stainless steels used in PWR operation • Type 316 CW Type 347 (Nb) Type 321 (Ti) Type 304• Rely more on IASCC database, and minimize the need to perform parametric IASCC studies


Strategies

Continue to develop deeper and broader collaboration with external R&D programs, in both US and international:

Collaboration with DOE-LWRS (Light Water Reactor Sustainability) Collaboration with EDF and MAI (Materials Aging Institute) New opportunities with DOE-NSUF (Nuclear Science User Facilities)

Increase R&D efforts to address materials issues in VVER and PHWR (CANDU)

Coordinate more effectively among the materials programs• New Materials Program Structure for better realignment

Collaboration, Coordination, & Integration


Member Engagement

Strong engagements with US and international members:• Improve relevancy through issue-based fundamental R&D, in order to achieve better

relignment and complement to issue program R&D • Increase R&D efforts to address materials issues in VVER and PHWR (CANDU)• Re-organization will improve the results

Active Participation by issue program chairs• BWRVIP, MRP, SGMP, WRTC, PWROG-Materials

More positive member feedback It is critically important to maintain strong engagements with members

Goal: Stronger Engagements With Members




Kevin Hacker, DominionNDE Chair

Steve Swilley, EPRINDE Program Manager

Materials APCWednesday, August 31, 2016

NDE ProgramQuick Summary



NDE Program Summary – Key Deliverables

Title Product ID Date

Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements

3002007626 30-Mar-2016

Nondestructive Evaluation: Industry Best Practices for Performing Reliable NDE - Implementation Guide

3002007329 29-Apr-2016

Nondestructive Evaluation: Procedure for Manual Phased Array Ultrasonic Testing of Weld Overlays - Procedure: EPRI-WOL-PA-1

3002008330 09-Jun-2016

Issue Management Table - Concrete 3002007779 18-Dec-2016

Remote Visual Testing Round-Robin Study 3002007793 18-Dec-2016


NDE Program Summary – Challenges Calvert Cliffs OE on pressurizer nozzle to safe end

weld – MSIP– NDE issue with data quality– Addressed by 2013 NIFG guidance– NRC questions on effectiveness of MSIP

RAI’s will be addressed to the utility

Project on CASS round robin to assess state of technology– Challenged to get vendors to participate– Working on alternative paths for diverse teams to review

data

Project on alternatives to IGSCC requalification

Project on identification and assessment of low-value examinations with high outage impacts


NDE Deviations for 2016 Deviating from “Needed” guidance in DM weld NDE guideline 300200091 Nine Mile Point Unit 2

– Utility attempted but could not execute an encoded examination for several welds (attempts to comply resulted in ~9 man-rem); instead performed a non-encoded phased array UT

– The NDE Program identified no technical issues with the deviation PSEG Hope Creek 1

– Revision to a previously submitted deviation; the only change is the date for the next schedule inspection

– NDE Program identified no technical issues with the deviation revision

NDE Program is reviewing the guideline to determine whether any of the operating experience warrants changes


Self-Assessment May 2016

The focused self-assessment identified one (1) strength, eight (8) areas for improvement, and five (5) enhancements

The 13 findings have been entered into EPRI’s corrective action program

All findings were of significance “Non-QA Low”

All findings were administrative in nature

The self-assessment report is available on NDE Cockpit




Dan Patten, FENOCWRTC Chair

Greg Frederick, EPRIWRTC Program Manager

Materials APCWednesday, August 31, 2016

WRTC ProgramQuick Summary



WRTC Program Summary - Outline

WRTC Strategic PlanKey 2015/2016 Products

– ASME Code Activities – Key Highlights– Materials and Welding Issues– Irradiated Material Weldability– Personnel Development and CBT Activities– Powder Metallurgy and Hard-facing Activities



WRTC Team – Who we areUnited States (TAC) 22 of 22 US Utility Organizations participate in

WRTC (all operating BWR and PWRs)International Participation (TAC) KHNP - Korea CEZ NPP- Czech Republic (2017) EDF/MAI - France KKL - Switzerland CANDU Owners Group (COG) – Canada MVM Hungarian Electric (Paks) – Hungary Electronuclear – BrazilNon Utility Memberships IHI Corporation – Japan Fluor Enterprises AZZ WSI LLC - Welding Services Inc. AREVA (2015) Westinghouse (2015) KAPL – Bechtel Marine Propulsion (2016) Rolls Royce (2016)

EPRI MAPC Director:• Kurt Edsinger, EPRIExecutive Sponsor:• Tom McCaffrey, EntergyWRTC Integration Committee:• Program & APC Chair: Dan Patten, FENOC• Program Vice-Chair: Joe Weicks, Entergy• IC Lead: John Sisk, Energy NW• IC Lead: Marc Hall, Dominion• IC Lead: Charles Bonan, Edf• IC Lead: RC Folley, APS


Welding & Repair Technology Center – Strategic Plan

As nuclear power producing facilities age, there is an increasing need for technology to provide effective solutions and to support life extension objectives

Focus on both tactical issues and strategic research– Provide a framework for identifying, prioritizing, and

tracking fabrication and repair related technology “gaps” Technical Advisory Committee (TAC), EPRI staff,

Integration Committee Facilities (metallurgical lab, welding lab, materials

labs, etc.) Collaboration – National Labs, Universities, Internal

– Lead R&D activities and technology development to supply the necessary “TOOL” to address current and future repair, fabrication, and mitigation issues

– 11 Research Focus Areas (RFA) established to address WRTC Core and Support areas:

RFA

InternationalResearch Focus Area 1 - Nickel-Base Filler Metal Weldability Solutions

Research Focus Area 2 - Irradiated Materials Welding Solutions

Research Focus Area 3 - Identify, Research, Develop, and Mature Advanced Welding Processes

Research Focus Area 4 - Optimize Joining, Fabrication, and Repair Processes. Stress Optimization

Research Focus Area 5 - Small Bore Piping Asset Management

1Research Focus Area 6 - Transfer & Promote Fabrication & Joining Technologies into Codes, Standards, & Regulations

Research Focus Area 7 - Buried Pipe Asset Management / Repair Solutions

Research Focus Area 8 - Repair Solutions for Structures -Containment and Fuel Pool Asset Management

1Research Focus Area 9 - Tactical Implementation of Repair Methods

1

Research Focus Area 10 - Document & Evaluate Operating Experience for Welding & Repair Programs. Training Actions and Topics (CBTs, workshops)

Research Focus Area 11 – Thermal spray, coatings, and Hardfacing Applications (including Powder Metallurgy)


ASME Code Activities – Key Highlights

Quick Summary


ASME Code Activities – Key Highlights Research Focus Area 6 (RFA 6): Transfer & Promote Fabrication

& Joining Technologies into Codes, Standards, & Regulations: This Research Focus Area addresses the role of WRTC in promoting

Code and Regulatory adoption of code cases and guidance documents that provide resolution for industry issues.

This area is broad, covering multiple design types and systems. WRTC Role: Establishing technical bases, reports status, expands

implementation and supports revisions of Code Cases. Typically RFA 6 is the highest priority Focus Area

Supports Section XI, IX, B31.1, AWS, Post Construction and NRC-Industry Interaction and Meeting


ASME Code Activities – Key Highlights Most recent deliverables in the RFA 6 (2015-2016)

– 3002007901 (2nd Quarter 2016), WRTC: Technical Basis and Residual Stress Studies to Support EWR Case N-847 Technical Report

– 3002005552 (2nd Quarter 2016), WRTC: Welding and Repair Technical Issues in ASME Codes and Standards – 2015 (Annual Update)

– 3002005542: (4th Quarter 2015), WRTC: Operational Leakage in ASME Class Systems; Operability, Evaluation, and Repair.

– 3002005539 (4th Quarter 2015), WRTC: Essential and Emergency Service Water Issues –Update

– 3002005537 (4th Quarter 2015), WRTC: Repairs to Leaking American Society of Mechanical Engineers (ASME) Class Systems – Update

– 3002005536 (4th Quarter 2015), WRTC: Shielded Metal Arc Temper Bead Welding

– 3002005518 (3rd Quarter 2015), WRTC: Excavate and Weld Repair Demonstration Mockup Results – Preliminary Report


ASME Code Issues Report (3002005552 –WRTC: Welding and Repair Technical Issues in ASME Codes and Standards – 2015 (Publication Date – May 2016)

Product is a overview of all WRTC engagements with Code and Regulatory Issues: – Introduction/Scope– Resolution of NRC Comments to Code Case N-766-1– Excavate and Weld Repair (EWR) Code Case N-847– Branch Connection Code Case N-853– Removal of Socket Weld Gap in B31.1– Weld Residual Stress– Comments to Draft Regulatory Guides 1.147 and 1.193– Upcoming Code Activities– Status of Code Cases– Appendices

ASME Code Meeting Summaries Tech Basis Documents for N-847 & NRC Comments N-853 Tech Basis Documents & NRC Comments Pin Brazing Code Case

ASME Code Activities – Section XI - Key Highlights

Welding and Repair Technology Center:

Welding and Repair Technical Issues in ASME

Codes and Standards - 2015


2016 PVP Papers to Support Code Activities‒ PVP2016-63769 – Technical Basis for Code Case N-847 – Excavate and Weld Repair (EWR) for SCC

MitigationSteve McCracken, Jonathan Tatman (EPRI); Pete Riccardella (SIA)

‒ PVP2016-63815 – 3D Residual Stress Simulation of an Excavate and Weld Repair MockupFrancis Ku, Pete Riccardella (SIA); Steve McCracken (EPRI)

‒ PVP2016-63197 – Residual Stress Mapping for an Excavate and Weld Repair MockupMitchell Olson, Adrian DeWald, Michael Hill (Hill Engineering); Steve McCracken (EPRI)

‒ PVP2016-64008 – Incorporating Peening into ASME Section XI Code Cases N-729 and N-770 for PWSCC Mitigation in Alloy 82/182/600 Locations

Dennis Weakland (Ironwood Consulting), Paul Crooker (EPRI), Glenn White (Dominion Engr)

‒ PVP2016-64032 – Deterministic Technical Basis for Re-Examination Interval of Every Second Refueling Outage for PWR Reactor Vessel Heads Operating at Tcold with Previously Detected PWSCC

Glenn White, Kevin Fuhr and Markus Burkardt (Dominion Engr); Craig Harrington (EPRI)

‒ PVP2016-63902 – Technical Basis for Code Case N-853 – A600 Branch Connection Weld Repair for SCC Mitigation

Dave Waskey (AREVA), Steve McCracken (EPRI)

‒ PVP2016-64007 – Applications of Welding to Repair Irradiated Reactor InternalsWayne Lunceford, Jon Tatman, Steve McCracken, Nathan Palm (EPRI); Eric Willis (PG&E)


ASME Code Activities – Section XI - Key Highlights WRTC Comments –Draft Regulatory Guides

DG-1296 R.G. 1.147 Rev. 18 and DG-1298 R.G. 1.193– Project Goal: Compile and submit, for Owner/Operator members of

WRTC, comments on the draft Regulatory Guides Draft Regulatory Guides published in the Federal Register March 2,

2016 Comment period ended May 16, 2016

• DG1296 Draft Reg. Guide 1.147 Revision 18, Code Cases Proposed to Be Approved for Use With Conditions

• DG1296 Draft Reg. Guide 1.147 revision 18, Code Cases That Have Been Superseded

• DG1298 Draft Reg. Guide 1.193, ASME Code Cases Not Approved for Use


ASME Section XI Code Activities – Key Highlights (Backup Slides)

New and Revised Code Cases of Importance– N-847 and N-770-5 Excavate and Weld Repair for SCC Mitigation

N-847 - Approved May 2016 by SC-XI on recirculation ballot (Record # 10-1845).

N-770-5 - Approved by SG-WCS (Record # 14-2233), currently at SC-XI for 1st consideration letter ballot.

– N-853 Branch Connection Modification for SCC Mitigation

– N-666-1 Weld Overlay for Class 1, 2 and 3 Socket Welded Connections (expansion)

Expand to include Ni-alloys and filler materials, potential for repair of other degradation (SCC, thermal, etc.), and expanding to additional configurations (elbows, threaded connections)

– N-839 Shielded Metal Arc Ambient Temper Bead

Support Dominion Virginia Power (Marc Hall) in qualifying welding procedures to N-839 guidelines

– N-661-3 Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service

– N-786-2 Sleeve Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping


ASME Section XI Code Activities – Key Highlights (Backup Slides)

New and Revised Code Cases of Importance– N-513-4 Temporary Acceptance of Flaws in Moderate

Energy Class 2 or 3 Piping Section XI…– N-789-2 Pad Reinforcement of Class 2 and 3 Moderate-

Energy Carbon Steel Piping– N-XXX Rolled Plate Reinforcement of Class 2 and 3

Atmospheric Storage Tanks Goal is presentation and request for letter ballot at SC XI in

November

– N-860 Inservice Inspection of Spent Nuclear Fuel Storage And Transportation Containment Systems ASME Task Group formed at the request of NRC to

establish Code rules for in-service inspection of dry storage systems for spent nuclear fuel

Subcommittee (repair and mitigation) formed under the ESCAP Program (EPRI Fuel and Chemistry) with objective to Demonstrate / validate techniques for mitigation and repair (coordinated with ASME)


ASME Code Activities – Section IX and B31.1 - Key Highlights (Backup Slides)

Section IX Activities– Key Changes to Section IX QW-290 – Approved for 2017

Temper Bead Welding Qualification• Modified Essential Variable Table QW-290.4• Eliminate QW-406.8 (Control of interpass temperature), Modify QW-

406.1 (preheat requirements), Modified QW-406.9 (elimination of layer by layer recording)

• Mandatory requirement for hardness or impact testing removed.• Goal: Create consistency for qualification of temper bead throughout

industry B31.1 Activities

– Elimination of gap requirement for socket welds Failed at Standards committee Code Case initiated for nuclear specific applications (low temperature)

– B31P - Pre-heat and Post-weld heat treatment standardization Ultimate goal is standardization in all codes


Materials and Welding Issues – Key Highlights

Quick Summary


Simple test couponMock up test

Alloy 52 Weldability Solutions• Background - Alloy 52 Alternative

• High-chromium nickel-base Alloys 52 and 52M are currently used for PWSCC mitigation repairs and new component fabrication

– Alloys 52 and 52M are plagued with poor weldability and are susceptible to microfissuring

• Current WRTC Activities1. Weldability Solutions – Alloy screening

• Simplified field deployable DDC screening test for high Cr alloys

• Standards for screening test2. New alternative high-chromium nickel-

base weld metal with:• Improved weldability and superior

resistance to microfissuring• Maintain PWSCC resistance, material

properties to alloy 690

Concept for Field Deployable Screening Test

Restraint : high Duration : high Cost : high

Restraint : med Duration : low Cost : low


Cast Pin Tear Test Results – 52M-Hf and 52M-Ta compositions (STR < 150C, Fraction eutectic > 2%)

0%

20%

40%

60%

80%

100%

0.625 0.75 0.875 1 1.125 1.25 1.375 1.5 1.625 1.75 1.875 2 2.125

Max

imu

m C

ircu

mfe

ren

tial

Cra

ckin

g

Pin Length (in)

52-Hf 52MSS-C FM82 (AB8573) FM82 (YB8908) 52-Ta

Good

Bad

Summary of Project Status - Alternative High-Chromium Nickel-base Weld Metal

• Complete‒ Computational modeling and DOE studies of

solidification behavior‒ Button melting and weldability experiments ‒ Manufacture of target 52M-Hf, 52M-Hf-Mo, 52M-Ta &

52M-Ta-Mo wires‒ Weldability testing and characterization of 52M-Hf &

52M-Hf-Mo wires• In Progress

‒ Strain-to-fracture (STF) testing of 52M-Ta & 52M-Ta-Mo wires at OSU

‒ Characterization of 52M-Ta & 52M-Ta-Mo wires• Near Future and N+2

‒ Dilution testing of 52M-Ta variants with CF8A 3rd quarter 2016

‒ Select optimized target heat 4th quarter 2016‒ Testing with various welding processes and

configurations‒ Full scale mockups & CGR testing Dual Wire Feeder


Irradiated Material Weldability

Quick Summary


Key 2016 Products & Challenges - Irradiated Material Weldability WRTC Research Focus Area (RFA 2): Irradiated Materials

Welding Solutions– Highly collaborative effort between EPRI (WRTC, LTO, MRP,

BWRVIP, NDE) and DOE– Current work is largely focused on:

Comprehensive understanding of the metallurgical effects of welding irradiated austenitic materials

Develop and validation of advanced welding processes Near-term solutions and guidance for irradiated material weld repair

– Project areas being addressed in a parallel path to support final goal of welding in hot cell

1. Development of an Effective heat input and dilution formula specifically for advance welding processes (backup slides)

2. ASME Code Activities and Development of weldability maps to highlight boundaries for application of various welding processes (backup slides)

3. Irradiated Material Development and Characterization Plan (backup slides)

4. Development of advanced welding methods for highly irradiated materials (backup slides)

5. Validation tests at ORNL to support advance welding process development (Hot Cell)

Advance Welding Development and

Modeling

Material Archive

Cubicle Fabrication

Validation Tests – Hot Cell

Code and Effective Heat input


Irradiated Material Weldability: ORNL Welding Facility – 2016 Status ORNL hot cell welding facility development in progress:

– Completed Milestones: Installation of Cubicle within Hot Cell Facility Electrical and Plumbing of friction stir weld (FSW) System Laser Safety Basis Report for DOE Review and Approval Cold Run of FSW System at Hot Cell Facility

– In-progress Milestones: Finalization of FSW Procedures Electrical and Plumbing of Laser Welding System DOE Approval of Laser Safety Basis Report Cold Run of Laser Welding System

Parallel Irradiated Welding Path: Recent DOE award involves collaboration between EPRI, Westinghouse, and Boise State to conduct laser welds and advanced characterization on representative 304 SS EBR-II irradiated materials

Huang, Y., JNM, 2015


Irradiated Material Weldability: ASME Section XI Activities (Backup Slide) ASME initiated an action (Record No. 10-1842) to review and evaluate

guidance and rules in Section XI when welding on irradiated materials– Review showed Section XI was inconsistent and needed comprehensive revision– Based on EPRI guidance (BWRVIP-97 R1 and MRP-379)

helium content and effective weld heat input in lieu of neutron fluence and theoretical heat input should be used as criterion for determining weldability

Current direction of ASME Working Group:– For code cases primarily applicable to RPV nozzle DM welds or upper / lower

head penetrations Remove existing criteria related to fluence Add a statement that the code case is not applicable to repair of reactor

internals– For code cases that may be applied to reactor internals

Remove existing criteria related to fluence Add criteria based on helium concentration similar to IWA-4661(f) and N-516-

4 Include language allowing generic disposition of components located

significantly above / below the core (< 0.01 appm He based on generic evaluation)

– For locations near the core, owners can apply the EPRI guidance Technical basis paper for suggested code changes published at 2016

PVP Conference– PVP2016-64007, “Applications of Welding to Repair Irradiated Reactor Internals”

0.01

0.1

1

10

0.1 1 10 100

Effe

ctiv

e H

eat I

nput

(KJ/

cm)

Helium Concentration (appm)

JOG-GTAW-BM - No Cracking JOG-GTAW-BM - CrackingJNES-GTAW-BM - No Cracking JNES-GTAW-BM - CrackingJNES-GTAW-WM - No Cracking JNES-GTAW-WM - CrackingLBW-BM - No Cracking LBW-BM - CrackingLBW-WM - No Cracking LBW-WM - CrackingLBW - Remelt Trials - No Cracking LBW - Remelt Trials - CrackingJNES-GTAW-WM - GBD LBW-BM - GBDLBW - Remelt Trials - GBD

*

*

*

*

*

*

* *

* - Denotes single pass result

*


Irradiated Material Weldability Testing – Deliverables and Milestones Deliverables

– EPRI Tech Report # 3002005553, WRTC: Advanced Laser Welding Technology for Irradiated Reactor Internals, issued March 2016

– EPRI Tech Report BWRVIP-97 Revision 1 , Guidelines for Performing Weld Repairs to Irradiated BWR Internals, issued December 2015

– EPRI Tech Report # 3002005545, WRTC: Advanced Welding Methods for Irradiated Materials (status of hot cell setup), issued December 2015. Status report update to be issued December 2016

– EPRI Tech Report # 3002005531: Heat input Efficiency Equation (Cracking Threshold, and screening of welding processes for highly susceptible materials.) Issued October 2015

End of 2016 – Mid-2017 Milestones:– Resolve irradiated material ASME and Regulatory Code wording issues– Finalize operating procedures (ORNL), and Phase 2 and 3 materials (HFIR)– Finalize “cold” welding operations within cubicle at ORNL hot cell facility– Finalize test matrix and acceptance criteria for welding on irradiated materials– Obtain DOE approval of cubicle safety evaluation (operation of welding cubicle)– Begin parallel path collaboration with Westinghouse Churchill Site and Boise State on 304

SS irradiated materials– Initiate welding trials on Phase 1 hot materials at ORNL

PWSCC Repaired via FSW

Laser weld with stress imaging


Personnel Development and CBT Activities

Quick Summary


WRTC - Personnel Development WRTC creates and sponsors development opportunities for members:

– Transferring knowledge is of high importance to WRTC members – Classes/workshops conducted at WRTC TAC meetings (June/December) and regionally at utility

locations– Presenters include EPRI staff, utility members and subject matter experts – Computer based training and webinars accompany traditional classes

Topic Where Provided Date / Number AttendedFundamentals of Welding •OPG: Whitby, Ontario (Assisted by York Chan) September 2015, ~30Operational Leakage Workshop •Palm Coast, FL

•New Orleans, LAJune 6, 2015 ~50December 2014 ~ 40

Introduction to Welding Programs –Emphasis on Welding Metallurgy

•Entergy - Pilgrim Nuclear Station (Mass, Cape Cod) March 31, 2015 ~ 25

Training for Repair/Replacement Engineers

•Prairie Island, MN•Entergy, Jackson, MS

January 13 – 15, 2015 ~ 25June 1, 2015 ~35

Introduction to Construction Codes • Wolf Creek, KS November 2014 ~ 32

Buried Pipe Workshop •Fort Myers, FL – Offered as part of Welding and Repair Conference

June 2014~55


WRTC will sponsor the following development opportunities in 2016:– Socket weld training was also developed in 2016 for craft personnel (PowerPoint) which is

intended to be used as a supplement to utility training This may be converted into a full CBT in the future

– 3002005542: (4th Quarter 2015), WRTC: Operational Leakage in ASME Class Systems; Operability, Evaluation, and Repair.

Topic Where Provided DateTraining for Repair/Replacement Engineers (ASME Section XI)

•Duke, Charlotte, NC•TVA, Chattanooga, TN•Xcel Energy, Monticello - MN

April 19 - 22, 2016August 1- 4, 2016August 29- 9/1 2016

Introduction Weld Residual Stress / Update on PWHT

June WRTC TAC Meeting, Portland, ME June 22 - 23, 2016

ASME Section IX, Overview and Practical Applications

Entergy, Jackson, MS August 2 – 4, 2016

Operational Leakage Webinars On-line Training September 15 and 29, 2016

Failure Analysis December WRTC TAC Meeting – Palm Coast, FL December 7-8, 2016

Completed

WRTC - Personnel Development


WRTC - Personnel Development (Backup Slide)

Computer Based Training 12 CBT modules completed using differing

techniques that maintain trainee attention and enhance the learning experience such as video clips, animation, audio voice over's, and quizzes were developed in 2015

Available for self study through WRTC Cockpit

CBT Module Topics:1. Nuclear Welding Program Overview2. Introduction to Welding, Brazing, and Fusing3. Introduction to Weld Configurations,

Considerations, and Defects4. Introduction to Materials and Welding

Metallurgy- Part 15. Introduction to Materials and Welding

Metallurgy - Part 26. Introduction to Welding Codes7. Introduction to ASME Section IX8. Weld Filler Material9. Welding Program Part 110.Welding Program Part 211.Alternate Repair Methods12.Introduction to Inspection Techniques


WRTC - Personnel DevelopmentDevelopment Activities and Workshops for 2017 is still being

determined, but some thoughts for 2017 are:– Weld training for Design Engineers– Performing impact testing for welding procedure qualifications– Regional classes– Webinars and CBTs

2017 International Welding Repair Conference in Orlando Florida– Generation and Nuclear Tracks (Handout)


Powder Metallurgy and Hardfacing Activities

Quick Summary


Powder Metallurgy – HIP (WRTC RFA 11)Valve Seats -- Cobalt-free alloys

Cobalt-free alloys (Backup Slides Available) Difficult to apply crack-free (NOREM 02) via welding Exhibit low-service temperature capabilities; <200C

Goals: Develop Co-free hardfacing alloy with galling and sliding wear resistance

from RT thru 350C (similar properties to Stellite 6) Optimize hardfacing alloy properties, w/o weldability issues

Results to Date Cobalt-Free NitroMaxx-PM Alloy applied via PM-HIP Valve seats can be joined via similar SS-to-SS weld or direct PM-HIP of

hardfacing to cast parts (eg. gate, globe) 3002008030 - Final Report on Development of NitroMaxx-PM (Q3-2016)

– NitroMaxx-PM: A Cobalt-free Stainless Steel Hardfacing Alloy for Valves and Other Nuclear Reactor Component Applications

Image: Courtesy of Weir Valves

19mm NitroMaxx-PM hardfacing applied to 12” diameter seat ring


NitroMaxx-PM Development

Nitromaxx-PM Properties Demonstrates excellent wear and galling properties at

343C (650F) comparable galling properties as Stellite 6 High fraction of carbides and nitrides Increased hardness, strain-hardening Similar surface deformation microstructures as Stellite 6

2016/2017 Objective

Implement product to support valve replacement activity with source term issues

Continue Cyclic-Loop Testing– Planned with EdF and Velan on 6” diameter valves

Continue Corrosion Testing– Assess under BWR and PWR environments at GRC

Initiate Large-Scale Galling Tests – Larger-scale galling tests planned w Battelle


WRTC Strategy and Strategic Challenges


WRTC Strategy & Strategic Challenges

– Challenges working in hot facilities (RFA 2)– Strategic Plan for TAC meeting organization for improved international engagement

ASME materials Repair Options and Regulatory approval

– Going into Base in 2017: More international members Continued engagement of vendors

– Nuclear Promise Risk informed Repairs, Case N-662-1 Identify best practices for repair and replacement activities toward reducing costs Implementation of welding critical factors for repair applications.

– IC membership changes New TVA, Keith Dietrich, Entergy, Joe Weicks to Vice Chair, New Charles Bonan,

EDF, John Sisk, Northwest Energy (thanks for support as past Vice Chair)




Keith Fruzzetti, Technical Executive

Materials Degradation and Aging Action Plan Committee Meeting

Wednesday, August 31, 2016

Potassium Hydroxide (KOH) as an alternative to Lithium

Hydroxide (Li-7)Qualification/Demonstration Plan



Li-7 for PWR Primary pH ControlWhy is pHT Control Important?Required by PWR Primary Chemistry Guidelines (≥ 7.0) for: materials, fuel reliability, and radiation fields

Reduces general

corrosion

Increases iron solubility

across the core

Stabilizes fuel crud

Impacts PWSCC initiation

Use of natural Li (~92% Li-7, ~8% Li-6) would greatly increase tritium production Need enriched Li-7 (≥ 99.99%)


Source: IAEA PRIS Database. Updated 29 Sept 2015

Li-7 for PWR Primary pH ControlWhy Investigate an Alternative? Some utilities were challenged / unable

to procure Li-7 in 2015– Production now back up in China and

Russia but price has increased significantly

– Dependability of current supply routes unclear

Operational considerations– Flex power ops GREATLY

increases Li-7 demand– Growing PWR fleet

Molten salt reactors would greatly increase demand (next slide)

GAO-13-716, “Managing Critical Isotopes: Stewardship of Lithium-7 Is Needed to Ensure a Stable Supply”, Sep. 2013.

Press Release, House Committee on Science, Space, & Technology, “GAO Raises Questions about Adequate Supply of Lithium-7 for Nuclear Power Reactors”, Oct 9, 2013.


Potassium Hydroxide (KOH): Motivation

ELIMINATE the Significant Vulnerability of Li-7 Supply

Approximately 26,500 kg LiOH•H2O / yr / unit needed*(Estimated average yearly use for a PWR: 35 kg**)

Molten Salt Reactor (LiF – BeF2 – ThF4 – UF4)

* Based on information from: Engel, J.R. et al., “Molten-Salt Reactors for Efficient Nuclear Fuel Utilization Without Plutonium Separation”, ORNL/TM-6413, Aug 1978. Basis: 1000 MWe.

** Includes lithium required to saturate a CVCS resin bed

(72% – 16% – 12% – 0.4%)

Eliminate dependence on Li-7 supply

– Existing supply chain vulnerability

– Advent of flexible operations

– Growing worldwide PWR fleet

– A single Molten Salt Reactor(1000 MWe) requires as muchLi as 760 commercial PWR units


Additional Benefits of KOH Lower Operational Costs

– LiOH•H2O: Approximately $2,500/kg– KOH: Approximately $25/kg

(Reagent Grade)– Standard vs Lithium saturated CVCS

bed (approx. $300 vs $6000 per ft3)– Estimated savings per year

Each PWR unit: $140kU.S. Fleet (65 PWR units): $9.1M

May be more beneficial for Fuel– Data indicates much lower corrosion rates

May mitigate IASCC* initiation (e.g. baffle-former bolts)– Much lower lithium concentrations possible with KOH

*IASCC: Irradiation Assisted Stress Corrosion Cracking

10

100

1000

10000

1 10 100 1000 10000 100000

Co

rro

sio

n R

ate

(mg/

dm

2)

Concentration of Cations (ppm)

Zircaloy 2

NaOH

LiOH

KOH

Cor

rosi

on R

ate

(mg/

dm2 )

Corrosion Rate of Zircaloy 2 at 360°C

Concentration of Cation (ppm)

H. Coriou, L. Grall, J. Neunier, M. Pelras, and H. Willermoz, “The Corrosion of Zircaloy in Various Alkaline Media at High Temperature”, Corrosion of Reactor Materials, Vol. II, 193, IAEA, Vienna (1962).

NaOH

LiOH

KOH


Two Paths to Maintaining pH Control Capabilities

Stay with Li-7 Qualify / Demonstrate KOH

Optimized Usage

Li-7 Recovery

Alternative Enrichment Processes

Stockpile

How long would stockpiles last, if usage is optimized? Is FULL qualification necessary if there is no supply?

Higher upfront research costs, Lower operational costsLower upfront research costs, Higher operational costs

Materials

Chemistry Control

Fuels

Radiation Safety


Feasibility of KOH vs LiOH for PWR Primary pH ControlPublished October 2015 (3002005408) – Reference/Early R&D

VVER Operating Experience– Successful use of KOH for

over 40 years– Generally low corrosion and

very low radiation fields– No observed Crud Induced

Power Shift (CIPS)

Materials

• Initiation and CGR of austenitic stainless steel and nickel based alloys

Fuels

• Corrosion and hydriding of zirconium fuel cladding – with crud and boiling

Chemistry

• Management of Li and K for pHT control

• High temperature chemistry

Radiation Safety & Radwaste

• Radiation Fields• Dose pathways• Waste classification• Effluents

Appears very promising. Some next steps underway. Detailed multi-year plan developed.

Key Gaps

Important differences between VVER and Western-PWRs– Materials: Titanium-stabilized SS (VVER) vs nickel-based alloys (PWR)– Fuel cladding: Both zirconium alloy (KOH less corrosive), but low crud

and lower boiling (VVER)– Chemistry: Ammonia for hydrogen (VVER) vs dissolved hydrogen gas

(PWR), Li/K new to PWRs– Worker dose & Radwaste: Potassium activation products (VVER)


Detailed Plan for Qualifying KOH Developed (Summary)

Fuel Vendor Assessment Experimental Loop Testing Fuel Exams

Crack Initiation & Crack Growth Rate Testing–Non-irradiated testing

Stainless Steel and Alloy 600–Irradiated testing

Stainless Steel

Activation species and dose pathways Effect on plant radiation fields Effluent and radioactive waste handling

High temperature chemistry (MULTEQ) Purity specifications Multiple alkali (Li & K) modeling and control

Materials Testing

Fuels Testing and Exams

Radiation Fields and Radwaste

Chemistry / pH Control

MRP

FRP

RS

Chem


Full Qualification / Demonstration Plan – Shortest Timeline

What is truly

necessary in the

face of no Li

availability?

“Plan B”

Phase 1: Qualification ahead of the PWR plant trial (3 years)Phase 2: PWR plant trial (3 cycles)

Start of Phase 2

“Plan A” (Reasonably Conservative)• Detailed project scopes developed for each technical item

• Phase 1: 3 years to plant trial (possible application at many units)• $5M

• Phase 2: Plant Trial• At one unit: $3 - $5M (fuel exams)


Next Step: Investigate a “Plan B”

Can we eliminate CGR Testing?

Can we reduce the time/scope of initiation testing?

Motivation/Scenario

• All Li-7 supply is gone.

• Operate plant with alternate pH control chemistry, or shutdown.

• What is the absolute minimum to have been completed to allow operation with KOH?

“Plan B” Effort

• Work directly with a utility willing to consider this premise.

• Include 3 – 5 utility experts

• Requires executive level input.

Can we eliminate these evaluations from the qualification process, and simply evaluate as part of the trial application?

Work with fuel vendors to define acceptable risks




David Steininger and Nathan PalmEPRI


Environmentally Assisted Fatigue

EPRI Overview



Content

BackgroundEPRI EAF PerspectiveEPRI Approach to EAFCurrent Activities

– Analytical– Experimental

Summary and Conclusions


Background Nuclear plants designed in accordance with ASME Section III are

required to address fatigue usage– Cumulative usage factor (CUF) must be less than 1.0– CUF is calculated using ASME Code S-N curves which are based on testing in

air Since 1999, plants applying for license renewal have been required to

address environmentally assisted fatigue (EAF)– Fatigue specimen testing in water environments has shown reduced cyclic life– Consideration of EAF has typically been accomplished through the application

of an environmental factor, Fen– Demonstrating that CUFen is less than 1.0 has been a significant challenge for

the industry High calculated CUFen values have not been substantiated by actual

plant service experience


Current EPRI Perspective on EAF

• No urgent need for improved methods; most plants have submitted applications and received NRC approval• Locations not meeting regulatory criteria using current analytical methods can be addressed by elastic plastic

analysis (EP) or monitoring (both expensive)

License Renewal

• Introduce an appropriate level of conservatism in analysis and test data that will likely be required for 80 years

Second License Renewal

• No significant issues with current licensing period• Introducing an appropriate level of conservatism will be useful for longer-term operation• Alternative EAF analysis methodologies have been submitted for consideration and need NRC approval

New Plants

• Some plants have used a reduced number of design cycles based on base-loading to satisfy cumulative usage requirements

• Load following operation may exceed cumulative usage factor requirements

Flexible Operations


Overview of EPRI EAF Effort

Objective

Ensure that EAF can be addressed in both current

& new plants in a consistent manner that meets nuclear safety

objectives while assuring appropriate level of

conservatism

Applicability

Better understanding leads to increased accuracy of

environmental fatigue curves and more accurate fatigue crack

growth rates that can then be used to optimize the fatigue

licensing basis.

Data are expected to be available to support some of the later license renewal applicants,

applicants for extended operation (60-80 years), flexible operation considerations, and new plant

applicants.

Results Implementation

Develop the technical basis for ASME Code modifications and

Code Cases that support refined procedures for assessing fatigue

environmental factors

Promote consistent procedures for use by vendors, construction firms,

and utilities

Support ASME Section III and XI Code revisions that explicitly incorporate EAF procedures


Environmentally Assisted Fatigue RoadmapEnvironmentally Assisted Fatigue Gap

Analysis and Roadmap for Future Research (Dec 2012, Report ID 1026724)

– Gap prioritization performed by industry expert panel

– 21 gaps identified as high priority– 7 hypotheses proposed to explain the

apparent discrepancy between test data and field experience


Environmentally Assisted FatigueSponsorship & Organization

EPRI Programs funding EAF Activities

Materials Reliability Program (MRP) Boiling Water Reactors Vessel Integrity Program (BWRVIP) Advanced Nuclear Technology (ANT) Primary Systems Corrosion Research (PSCR)

Overall Coordination - David Steininger ([email protected])

‘Short-Term’ or Analytical Committee led by Nathan Palm of BWRVIP ([email protected]) Address knowledge gaps pertaining to conservatisms in analytical methodologies–Develop the technical basis for ASME Code modifications and Code Cases that support refined –procedures for assessing fatigue environmental factors

‘Long-Term’ or Experimental Committee led by David Steininger ([email protected])Address knowledge gaps pertaining to mechanistic understanding of EAF–Resolve perceived discrepancies between EAF methodology, existing test data, and industry operating –experienceChair, International EAF committee coordinating EAF testing world wide. Committee spear heading –testing of full, prototypical test fixture.





Analytical Committee ActivitiesExisting EPRI Guidance Documents

Selection of Locations for EAF Evaluation

• Environmentally-Assisted Fatigue Screening, Process, and Technical Basis for Identifying EAF Limiting Locations• Issued September

2012• Product ID: 1024995

Calculate CUF

• Stress-Based Fatigue Monitoring Methodology for Fatigue Monitoring of Class 1 Nuclear Components in a Reactor Water Environment (FatiguePro Basis)• Issued December

2011• Product ID: 1022876

Calculate Fen

• NUREG/CR-6909• Regulatory Guide 1.207• Both documents

currently under revision

Calculate CUFen

• Evaluation of Controlling Transient Ramp Times Using Piping Methodologies When Considering Environmental Fatigue• Issued September 2007• Product ID: 1015014

• Guidelines for Addressing Environmental Effects in Fatigue Usage Calculations• Issued December 2012• Product ID: 1025823


Analytical Committee ActivitiesEfforts directed by EPRI EAF Focus Group comprised by

industry fatigue practitionersTwo projects to propose changes to conventional fatigue CUF

calculations are underway– Alternative Approaches for ASME Code Simplified Elastic Plastic

Analysis– Fatigue Usage Gradient and Life Factors

Proposed changes to CUF calculations would partially offset Fenpenalty under EAF conditions

Proposals being made to and vetted by appropriate ASME Code committees– WG Design Methodology has jurisdiction for applicable Code sections– WG Environmental Fatigue Evaluation Methods also being engaged for

additional stakeholder input


Experimental Committee Activities

Begin with high-priority gaps from EPRI EAF RoadmapUnderstand relationship of gaps to hypotheses

– Understand and characterize critical environmental effect variables– Reconcile lab data and operating experience

Use of RFP process to identify industry capabilities and solicit input

Establish a 5 year collaborative & coordinated testing plan


Analytical Committee Activities


Alternative Approaches for ASME Code Simplified Elastic Plastic Analysis - Background ASME Code simplified elastic-plastic analysis (application of the Ke

factor) is recognized as one of the largest sources of conservatism in fatigue analysis

Elastic-plastic analysis may be performed in lieu of the use of the ASME simplified rules– Expensive to implement– No defined rules or criteria for acceptable elastic-plastic analysis

Past attempts have been made to propose alternative rules have been “unsuccessful”– Complicated to implement– Contain discontinuities in the solutions– Require new stress analyses– Not endorsed by the NRC


Alternative Approaches for ASME Code Simplified Elastic Plastic Analysis – Project Approach

Evaluate simplified elastic-plastic rules in other Codes (RCC-M, JSME, etc)

Address recommendations made by the Welding Research Council in Bulletin WRC-361

Develop a new Code proposal that:– Can be shown to be conservative relative to elastic-plastic analysis– Requires no new stress analysis– Covers common structural materials- austenitic stainless steel, nickel

based alloys, carbon steel and low alloy steel– Has the potential to reduce CUF values– Will in most cases offset the need for elastic-plastic analyses


Alternative Approaches for ASME Code Simplified Elastic Plastic Analysis – Project Status Proposed Code Case has been

developed Code Case methods have been

compared to elastic-plastic FEA results– Multiple cases considered– Proposed revision to Ke (Ke*) bounds

elastic plastic results Proposed Code Case presented to

ASME Code committees at August 2016 meetings


Fatigue Usage Gradient and Life Factors – Background Under ASME Code fatigue usage

calculations rules:– Allowable fatigue life is based on fatigue testing

of small diameter specimens and is subsequently applied to all components regardless of their actual thickness

– All component cyclic stresses are treated as uniform through-thickness membrane stresses and do not consider the presence of actual through-thickness stress gradients.

Fatigue life consists of two stages:– Formation of microcracks and growth of these cracks

to mechanical cracks– Growth of mechanical cracks to failure / load drop


Fatigue Usage Gradient and Life Factors – Approach The Gradient Factor (GF) accounts for the increase fatigue life

associated with through thickness stress gradients– Fatigue usage in plant components are primarily driven by high peak thermal

transient stresses at the inside surfaces and significant through thickness stress gradients

– Crack driving force will decrease as the crack grows through the pipe wall– Longer fatigue life results when the gradient stress is used rather than when

the peak stress is applied uniformly across the thickness The Life Factor (LF) accounts for increased fatigue life associated with

component thicknesses greater than the small diameter of fatigue test specimens

CUF values can be multiplied by the GF and LF to result in measurable reductions in estimated fatigue usage (especially in thicker piping)


Fatigue Usage Gradient and Life Factors – Status

Gradient and Life Factors have been developed for a sample problem

Additional calculations are needed to determine GF and LF values for a range of applied loadings, geometries, and materials

An ASME Task Group is being formed to provide peer review and determine means for incorporation of GF and LF into ASME Code rules.


Experimental Committee Activities


Background and Experimental Objectives “Checks and Balances” using field experience must be a consideration

– Field experience is not consistent with results of EAF testing using typical fatigue type tubular specimens undergoing classical loading under PWR environmental conditions NUREG/CR-6909 testing was expertly performed and consistent with

classical type loading conditions used for fatigue evaluation of materials Use “Separate Effects” test data to identify operational variables that

affect fatigue life not represented in previous testing– Test fixtures similar to that used in NUREG/CR-6909 testing, but capable of

variable loading and environmental conditions Identified operational variable transients that increase fatigue life relative

to NUREG/CR-6909 results will be validated against a full, prototypical test


Prototypical Test International EAF collaborative group supports test

– EDF– Rolls Royce– AMEC– AREVA– EPRI

Test fixture possibility– 4” nozzle with dissimilar metal weld to stainless steel

Effort must be co-funded RFP being developed

– Completed by July 15th

– List of contractors available Effort begins in 4th quarter of 2017

– Start time dictated by yearly funding planning


Significant Results from “Separate Effects” Tests

Simulating actual plant transients involving load and temperature significantly lowers EAF effect

In these tests, no measurable effect on EAF of strain holding (simulating plant startup condition) was found– Consistent with other Japanese data– Not consistent with some European data

Testing is on-going under new contract– Further testing to investigate reason for favorable fatigue life for

typical plant thermal shock piping transientTheory: Strain effect in compressive strain zone is nullified


MHI S-N Testing

τ1

Strain(%)

0

100

Tem(℃)

325

τ3τ2

εa

Constant(I1-3,I1-4)In phase(NI2-3)

Out of phase(NI2-4)

τ1 τ3τ2

εa

Time

Fast

Fast

Fast

Fast

No Temp. Strain amp.εa

Temp. change Pattern

Positivestrain rate

change Pattern

Period (sec) Strain rate (%/sec）

(℃) (%) τ1 τ2 τ3 ሶ𝜀1 ሶ𝜀2 ሶ𝜀3I1-1 100 0.60 Constant Slow-fast 600 480 1000 +0.001 -0.0025 +0.0006I1-2 325 0.60 Constant Slow-fast 600 480 1000 +0.001 -0.0025 +0.0006I1-3 100 0.60 Constant Fast-fast 600 480 1000 -0.001 +0.0025 -0.0006I1-4 325 0.60 Constant Fast-fast 600 480 1000 -0.001 +0.0025 -0.0006

NI2-1 100-325 0.60 Out of phase Slow-fast 600 480 1000 +0.001 -0.0025 +0.0006NI2-2 100-325 0.60 In-phase Slow-fast 600 480 1000 +0.001 -0.0025 +0.0006NI2-3 100-325 0.60 In-phase Fast-fast 600 480 1000 -0.001 +0.0025 -0.0006NI2-4 100-325 0.60 Out of phase Fast-fast 600 480 1000 -0.001 +0.0025 -0.0006

Fatigue test matrix

Out of phase

In phase

Note) “Strain rate change pattern” does not account negative strain slope.

τ1

Strain(%)

0

100

Temp.(℃)

325

τ3τ2

εa

Constant(I1-1,I1-2)Out of phase(NI2-1)

In phase(NI2-2)

τ1 τ3τ2

εa

Time

Slow

FastFast

Slow


S-N Testing ResultsTest results: Fatigue Life

100

1000

10000

I1-1 I1-2 I1-3 I1-4 NI2-1 NI2-2 NI2-3 NI2-4

Isothermal Condition Non-Isothermal Condition

Fatig

ue L

ife in

PW

R e

nviro

nmen

t

Prediction by NUREG/CR-6909 Rev.1Prediction by JSME S NF1-2009Experimental Fatigue life

100ºCSlow-Fast

325ºCFast-Fast325ºC

Slow-Fast

100ºCFast-Fast

Out of PhaseSlow-Fast

Out of PhaseFast-Fast

In PhaseSlow-Fast

In PhaseFast-Fast


S-N Testing with Complex Transient LoadingTest Results

We cannot see significant effect of strain holding from those test results.


Review of “Separate Effects” Tests Results - KHNP/KAIST

Results of testing on hold times ( 60 and 300 seconds) shows mixed results

MHI hold time was up to – 1000 hoursRegardless of the strain rates, the fatigue life of 60 and 300 seconds strain holding condition are slightly longer than the model presented in NUREG/CR-6909, Rev1.

Overall, there is no clear effect of strain holding (unacceptable –scatter band) on EAF life for 316 stainless steel in PWR water.

Testing is on -going under new contractTesting to focus on zinc water chemistry–


NRC Interaction

Meeting held with NRC on June 30, 2016Presentations made on analytical and experimental activitiesFavorable feedback obtainedNRC is interested in periodic interaction on EAF topics


Summary and Conclusions EAF has mostly been addressed for plant operation up to 60 years, although

this has often required detailed analyses or transient monitoring Additional challenges may be encountered for plants to operate to 80 years or

perform flexible operation Development of modified analysis methods would help to reduce CUFen and

address these challenges Additional testing is needed to understand the disconnect between fatigue

specimen testing, plant component operating experience, and actural plant component dimensions

Testing that is more representative of actual plant operation is expected to provide data for future revision of the Fen factor

NRC wants EAF validation of any new data that is intended to modify NUREG/CR-6909 content– Will accept only test data validated by full prototypical component test

