annual nrc and industry materials technical exchange meeting … · 2016-08-10 · 2016 materials...

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

Annual NRC and Industry Materials

Technical Exchange Meeting –

August 4, 2016Chicago

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

Agenda

IntroductionsAction Item ReviewReview of the executive meetingNormal OE/status reports from the Issue Groups NRC discussion topics

– Baffle former bolt– MSIP indications (Calvert OE)– NRC Peening SE status– NRC BTP 5-3 status– NRC budget and restructuring status and impacts on materials programs

Discuss the industry screening process for deciding when to submit documents Industry describe impacts of NRC work stoppage on Appendix G General discussionPublic commentAdjourn


Drew Odell (Exelon)BWRVIP Integration Committee Chair

NRC / Materials Issue Programs Tech Exchange Meeting

International Materials Reliability Conference

August 4, 2016

BWR Vessel & Internals Project

(BWRVIP)


Overview Outline

• Current Members and Organization

• 2016 BWRVIP Major Tasks

• 2016 BWR Operating Experience

• Status of Key Topics with NRC

• Contact Information


Current Members and Organization


2015 BWRVIP Member Utilities

U. S. • DTE Energy• Duke Energy

• Energy Northwest• Entergy• Exelon

• FirstEnergy• NextEra Energy

• NPPD• PPL

• PSEG Nuclear• Southern Nuclear Company• Tennessee Valley Authority

• Xcel Energy

Intl• BKW FMB Energie AG – Switzerland

• Chubu Electric Power Company – Japan• Chugoku Electric Power Company – Japan• Comision Federal de Electricidad – Mexico

• Forsmarks Kraftgrupp AB – Sweden• Horizon - UK

• Iberdrola Generation – Spain• JAPC – Japan

• Kernkraftwerk Leibstadt – Switzerland• Nuclenor – Spain

• OKG Aktiebolag – Sweden• Ringhals AB – Sweden

• Taiwan Power Company – Taiwan• Tokyo Electric Power Company – Japan


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)

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)


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 activitiesBWRVIP SLR activitiesExtension of ISP for Subsequent License Renewal (SLR)



Develop new report screening criteria–Delivering the Nuclear Promise InitiativeScreens NEI 03-08 reports related to need for NRC review and approvalDraft screening criteria is currently under reviewExpected MAPC (Materials Action Plan Committee) endorsement August 2016Work with the staff following MAPC endorsement


2016 BWR Operating Experience



BWR 4 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



BWR 4 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


Status of Key NRC Topics


Status of Key Topics with NRC

The BWRVIP appreciates the staff issuing the following: – BWRVIP-18, Revision 2, Core Spray I&E Guidelines

Safety Evaluation– BWRVIP-183, Top Guide Grid Beam I&E Guidelines

Safety Evaluation– Hope Creek’s surveillance capsule (ISP) extension letter

to PSEG – BWRVIP-234, Thermal Aging & Neutron Embrittlement of

CASS Safety Evaluation


Status of Key Topics with NRC

The BWRVIP is anticipating the release of the updated “FAVOR” code for several PFM (Probabilistic Fracture Mechanics) related activities.


BWRVIP Key Contact Information

BWRVIP Executive ChairmanTim Hanley, Exelon– Phone: 630-657-3724– E-Mail: [email protected]

Utility Technical ChairmanDrew Odell, Exelon – Phone: 610-212-1155 – Email: [email protected]

EPRI Program Manager Andy McGehee – Phone: 704-502-6440 – Email: [email protected]


Together…Shaping the Future of Electricity

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

2016 Materials Programs Technical Information Exchange Meeting

PWROG Materials Committee UpdateHeather Malikowski, Chair PWROG MSC (Exelon)

Chicago, IL


2016 Materials Programs Technical Information Exchange MeetingPWR Owners Group MSC Agenda

• Significant Activities Since Last Information Exchange Meeting June 2015

• Areas of Coordination & Strategic Planning with EPRI MRP

• PWROG MSC Focus Areas for 2016/2017 • Future PWROG MSC Meetings • MSC PWROG Core/Planning Team Organization and Key

Contacts

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2016 Materials Programs Technical Information Exchange MeetingSignificant Activities Since Last Meeting June 2015

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. The dialogue with the NRC was favorable (slide 10).

• Issued final PWROG Report PWROG-15105-NP, Revision 0, “PWR RV Internals Cold-Work Assessment” (May 2016). The NRC made a formal request for the report for information only (slide 10).

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2016 Materials Programs Technical Information Exchange MeetingSignificant Activities Since Last Meeting June 2015

Generic Projects – Cont’d• Issued draft 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 (slide 11).

• Issued draft PWROG 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.

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

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2016 Materials Programs Information Exchange MeetingAreas of Coordination & Strategic Planning with EPRI MRP

• Reactor Vessel Integrity– Participating in ASTM E10.02 to remove conservatism from sigma term of

consensus ETC. – 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). Meeting held on March 2, 2016 to present the overall plan.

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2016 Materials Programs Technical Information Exchange MeetingAreas of Coordination & Strategic Planning with EPRI MRP- Cont’d• Reactor Internals

– Coordinating the industry response to the recent baffle-former bolt Inspection findings (slide 7).

– 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 steel.– Working with MRP and NRC on a statistical approach for

assessing CASS material in PWR reactor internals. Face to face meeting held on May 5, 2016.

• 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.

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2016 Materials Programs Technical Information Exchange MeetingAreas of Coordination & Strategic Planning with EPRI MRP- Cont’d• Baffle-Former Bolt Inspection Findings

– During the Spring 2016 outage seasons, owners have identified broken and degraded baffle-former-bolting (BFB) at Indian Point Unit 2 and Salem Unit 1. The anomalies are consistent with prior experiences at DC Cook Unit 2.

– Baffle-Former Bolt Focus Group Formed• Chair – Tim Wells – SNOC. First meeting held on May 19th in Cranberry

– Six Strategic Focus Areas (FA) of Consideration• FA #1: Cause/Extent of Condition / Interim Guidance / Regulatory Technical Interface• FA #2: Plant/Fleet Operating Experience Assessment (PWROG lead)• FA #3: Repair/Replacement (PWROG lead)• FA #4: Inspection/NDE• FA #5: Irradiated Testing Support• FA #6: Aging Management Assessment

– Near Term Needs• Guidance to plants entering the fall outage season• Understanding OE/trends in BFB data• Tooling issues/bolt availability

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2016 Materials Programs Technical Information Exchange MeetingPWROG MSC Focus Areas for 2016/2017

• Ongoing Programs to Support MRP-227-A - Reactor Internals PA-MSC-1473 – Baffle-Former Bolt Technical Support for the Fleet PA-MSC-1388 - Reactor Vessel Internals Industry Coordination PA-MSC-1299 – Guide Card Wear RAI Support PA-MSC-1288 – PWR Materials Assessment PA-MSC-1286 - Evaluation of Potential Wear: Thermal Sleeve Flange 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 WCAP-17096 report)

26


2016 Materials Programs Technical Information Exchange MeetingPWROG MSC Focus Areas for 2016/2017 – Cont’d

• Ongoing 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-1123 - Reactor Vessel Integrity Industry Coordination and NRC Interaction PA-MSC-1091 – Demonstrate Excessive Appendix G Margins for PWR RPV Nozzles PA-MSC-0938 – Update of Surveillance Capsule Fluence Summary Report WCAP-14044

• Other Ongoing Programs 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-1182 - Revision to BAW-1543 for Master Integrated Reactor Vessel Program

27



MSC-1288 “PWR Materials Assessment”• The PWROG developed a statistical methodology for determining and assessing material or

fabrication factors for PWR reactor internals components. In addition, as part of the revision 1 work scope the PWROG developed a generically applicable position to respond to the NRC’s concern about the potential presence of non-fastener components with greater than 20% cold-work.

– Transmitted final Report PWROG-15032-NP, Revision 0, “Statistical Assessment of PWR RV Internals CASS Materials” to the NRC.

• The PWROG met with the NRC on May 5, 2016. As part of the informal submittal the NRC issued five draft RAIs. After the meeting with the NRC on the 5th of May three of the RAIs were likely resolved.

– Transmitted final PWROG Report PWROG-15105-NP, Revision 0, “PWR RV Internals Cold-Work Assessment” (June 2016) to the NRC.

• Plan to meet with the NRC at a future date.• The overall goal of the submittals is to eliminate MRP 227-A Applicant /Licensee Action Item

(A/LAI) # 7, “Plant Specific Evaluations of CASS,” and help utilities address the question associated with A/LAI # 1, “Applicability of FMECA and Functionality Assumptions,” relative to cold work.

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MSC-1319 “Transitioning RV Integrity to Direct Fracture Toughness, Phase 1”

• Phase 1 of the overall program (this PA is Phase 1) is to assess the viability of implementing the three phase plan laid out in PWROG-15089, “Plan for Transitioning RV Integrity to Direct Fracture Toughness,” develop the detailed test matrix, and test reactor irradiation specification.

• Meeting held with the NRC in March 2016.

– The technical approach was well received with some technical items raised which the industry needs to ensure are adequately addressed. It is clear that the regulatory/licensing aspects of implementing this plan need to be examined before proceeding.

– Phase 2 to be brought to the PWROG MSC for consideration in August 2016.

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2016 Materials Programs Technical Information Exchange MeetingFuture PWROG MSC Meetings (2016/2017)

30

April 24-27, 2017Pittsburgh, PA (Joint)

December 5-8, 2016Marco Island, Florida (Joint)

August 29-31, 2016Las Vegas, NV

August 14-17, 2017Baltimore, MD (Joint)


2016 Materials Programs Technical Information Exchange MeetingMSC Core/Planning Team Organization & Key Contacts

31

Materials CommitteeHeather Malikowski, Exelon (Chair)

(610) 765-5864

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

Reactor Internals Industry Planning Team

Glenn Gardner(860) 440-0373

Reactor Vessel Integrity Core Team

Jim Nurrenbern, Ameren(314) 225-1908

Pressure Boundary Core Team

Steve Petro, AEP(269) 697-5048

Jim MolkenthinPWROG PMO(860) 731-6727

Tammy NatourAREVA

(434) 832-2763

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 P32

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.

Global Expertise • One Voicewww.pwrog.com


Bernie Rudell, Exelon, MRP IC ChairAnne Demma, EPRI, MRP Program Manager

Industry/ NRC Technical Exchange MeetingChicago

August 4, 2016

Materials Reliability Program (MRP)


MRP Mission

Significant materials issues in the late 1990s led to the formation of the Materials Reliability Program (MRP)

The objective of the MRP is to resolve existing and emerging PWR materials performance, safety, reliability, operational and regulatory issues


MRP Research Focus Areas

RFA Description

1 Reactor Vessel Internals Assessment, Modeling and Inspection

2 Reactor Vessel Internals Irradiated Materials Testing

3 Alloy 600/690 Management, Mitigation, and Inspection

4 Reliability of CASS Pressure Boundary Components

5 Pipe Rupture Probability Assessment

6 Stainless Steel Degradation Mechanism Studies

7 Fatigue Management (Thermal and Vibration Fatigue)

8 Replacement Materials Testing (Alloy 690/52/152)

9 Reactor Pressure Vessel Integrity

10 Environmentally Assisted Fatigue


2016 MRP Deliverables (1 of 2)

Product ID TitleItem Type

Planned CompletionDate Status

3002007392 MRP: Topical Report for PWSCC Mitigation by Surface Stress Improvement (MRP-335 Rev. 3) Report 2/29/2016 Completed

3002008359 MRP: Effects of Surface Peening on the Inspectability of Nondestructive Evaluation Report 4/15/2016 Completed

3002007383 MRP: Basis for ASME Section XI Code Case N-838—Flaw Tolerance Evaluation of Cast Austenitic Stainless Steel (CASS) Piping Components (MRP-362 Rev. 1) Report 4/30/2016 Completed

3002008636 MRP: Development of Probability of Detection Curves for UT of Dissimilar Metal Welds (MRP-262, Rev. 2) Report 5/2/2016 Completed

3002008082 MRP: Effect of Lithium Concentration on IASCC Initiation in Irradiated Stainless Steel (MRP-413) Report 6/30/2016 Completed

3002008084 MRP: Specification Guideline for PWSCC Mitigation by Surface Stress Improvement (MRP-336, Rev. 1) Report 6/30/2016 Completed

3002007897 MRP: Revised Technology for Reactor Vessel J-groove Weld Surface Examination (MRP-410) Report 7/1/2016 Completed

3002007934 MRP: Administrative Procedures (MRP-130, Rev. 4) Report 7/22/2016 Completed

3002007851 MRP: Summary of JSME Thermal Fatigue Assessment Guideline and Comparison with MRP Management Guideline (MRP-408) Report 9/5/2016 Completed

3002007852 MRP: Benchmark of Thermal Fatigue Management in France (MRP-409) Report 9/5/2016 Completed


2016 MRP Deliverables (2 of 2)

Product ID TitleItem Type

Planned CompletionDate Status

3002007850 MRP: Environmentally Assisted Fatigue Testing of Stainless Steel Under Non-isothermal and Complex Loadings (MRP-407) Report 9/30/2016 On

Schedule

3002007964 MRP: PWR Supplemental Surveillance Program (PSSP) Capsule Fabrication Report (MRP-412) Report 9/30/2016 On Schedule

3002008083 MRP: Basis for PWSCC Mitigation by Surface Stress Improvement (MRP-267, Rev. 2) Report 9/30/2016 Completed

3002007853 MRP: Management of Thermal Fatigue in Normally Stagnant Non-Isolable RCS Branch Lines (MRP-146, Rev. 2) Report 10/30/2016 On

Schedule

3002007960 MRP: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion (MRP-191, Rev. 1) Report 10/31/2016 On

Schedule

3002007849 MRP: PWSCC in Alloys 600 and 690: Quantitative Assessment of Crack Initiation Time (MRP-406) Update 11/18/2016 Completed

3002007948 MRP: PWR Bottom Mounted Nozzle Exam Zone Definition & Basis Development (MRP-411) Report 11/18/2016 On Schedule

3002007955 MRP: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internal Components (MRP-232, Rev. 2) Report 12/16/2016 On

Schedule

3002007933 MRP: Inspection Data Survey Report (MRP-219, Rev. 12) Report 12/18/2016 On Schedule

3002007951 MRP: RPV Integrity Primer (MRP-278, Rev.1), A Primer on Theory and Applications Report 12/18/2016 On Schedule


PWR Materials Operating Experiences in Past Year

2015– Cold Leg Instrument Nozzle evidence of potential leakage

> Later determined to not to be an RCS leak. 2016

– Axial flaw in DM weld at 4” Pressurizer Safety Nozzle, MSIP mitigated in 2006, was determined to be deeper. Root cause concluded flaw depth was miss sized in prior exams. Repaired by Full Structural Weld Overlay

> NDE Alert letter issued– BFB Inspection results in high % failed in two Westinghouse 4-Loop down flow units

> An Industry BFB FG has been established with MRP and PWROG participation– Thermal Fatigue Exams reveal flaws in two units

> Not an “emergent issue” as addressed in recently issued MRP IG - Thermal Fatigue FG has the lead on follow assessment of these issues

– Evidence of leakage at a Cold Leg RTD Instrument Nozzle> ET revealed an axial PWSCC flaw in nozzle, opposite j-groove weld. Repaired RTD nozzle with a Mechanical Nozzle Clamp per Code Case.

– RV Head CRDM/Nozzle Exams identified two shallow off axis indications just below the toe of the J-groove on one nozzle – repaired by grinding to PT white results

> Not an “emergent issue”.– ISI identified an acceptable by evaluation flaw in a 14” RHR return line off the hot leg

> Final Owner causal determination to be reviewed for disposition


Potential Non-Conservatism in BTP 5-3

MRP/BWRVIP and PWROG evaluated potential non-conservatisms in BTP– MRP/BWRVIP evaluated the conservatism of the BTP methods and

assessed safety significance of non-conservatisms– PWROG’s “Material Orientation Offset Approach” (MOTA) showed

existing P-T curve methods for PWRs have available conservative margin to offset BTP in most cases

Evaluation found some BTP methods are potentially non-conservative, depending on product formPerformed probabilistic fracture mechanics (PFM) analyses to

assess risk of continued use of existing BTP through 60 years– Risk to vessel integrity is negligible; no safety benefit to be gained from

revising vessel P-T limits to address BTP non-conservatisms– One exception: 1 BWR used BTP B1.1(4) for a beltline nozzle, the

impact of which cannot be assessed by PFM, and will likely require plant-specific effort to address

Awaiting final disposition from the NRC


Regulatory Interface

Received a fee waiver for NRC’s review of MRP-227-Rev.1, PWR Internals Inspection and Evaluation Guidelines Development work for MRP-227 Rev 2 to address GALL SLR has

established an acceptable timeline with NRC to maintain generic reference of MRP-227-A in GALL SLRMRP-335, R3, the technical basis for optimized inspection

intervals for Alloy 600 locations based on PWSCC mitigation by peening was submitted to the NRC for a safety evaluation (SE)– In parallel, peening mitigation has been incorporated into Section XI of

the ASME Code by revisions to the code cases for DMWs and RPVHPNs. The final code case is expected to be approved in August 2016.

– The initial implementation of peening in the US PWR fleet was successfully completed in spring 2016. Another plant will peen reactor coolant inlet/outlet and BMN nozzles in the fall of 2016. Additional peening mitigation projects at PWR plants in the US are under contract.


Tim HardinTechnical Executive, MRP, EPRI

Industry/ NRC Technical Exchange Meeting

Chicago

August 4, 2016

Status of ASME Appendix G

Small Surface Flaw Issue


Background In August 2012, NRC informed industry of a concern regarding

potentially high Conditional Probability of Failure (CPF) for RPV when a small inside-surface circumferentially-oriented flaw is postulated during cooldown2012-2013 MRP performed independent FAVOR analyses (MRP-

368)2014: MRP work continued, seeking to define a potential solution

to ensure safety goals (CPF < 1E-6) are met under all conditions2014: BWRVIP began investigation of impact of shallow surface

flaws on BWR leak test conditionsBefore work could be completed, computational anomalies in the

FAVOR software were identifiedMRP informed NRC in August 2014NRC and ORNL have reported efforts to revise FAVOR; new

version expected soon


Stress Free Temperature Approach

FAVOR uses a Stress-Free Temperature (SFT) approach for handling cladding residual stress– At SFT, RPV shell is assumed free from

thermal strain; stress due to differences in Coefficient of Thermal Expansion (CTE) between clad and base metal, and clad residual stresses, offset each other

Derivation of the recommended Stress Free Temperature (SFT) used in FAVOR was originally detailed in NUREG/CP-0166 (ML042230476) – Measured strains were resolved into

circumferential and longitudinal stress components. The circumferential stress component was larger and was used to derive SFT = 488°F

Figure from ORNL/TM-2015/59531/REV-01

(ML16043A170)


SFT for a Circumferential Flaw

To date, both the NRC and industry analyses have used SFT = 488°F (based on the circumferential stress component, which acts on an axial flaw) However, the small surface flaw issue is a circumferential flaw issue

– “All inner-surface flaws are oriented circumferentially because in the examinations of the PVRUF vessel all of the observed flaws in the cladding were circumferentially oriented. This observation was consistent with expectations, because weld-deposited cladding is applied to vessel inner surfaces as a series of circumferential weld passes. In the FAVOR model, all inner surface-breaking flaws are associated only with the vessel cladding process.” (from ORNL/TM-2015/59531/REV-01)

In 2014 draft report that reported the FAVOR anomalies, EPRI asked, What is the SFT for a circumferential flaw? ORNL presentation at January 19, 2016 public meeting provided the answer

(“Impact of using stress free temperature of 364 F on Shallow Flaw Issue,” by Terry Dickson and Richard Bass (ML16021A008))– “PFM Results: SFT of 364 reduces CPI for 0.03t flaw by factor ~30 and

CPF by factor ~400”


Plans for Future Work on Small Surface Flaw Issue

NRC has also reported work on an alternate cladding KI model that uses cladding residual stress data obtained as part of the international cooperative program Network for Evaluation of Structural Components (NESC)NRC/ORNL paper PVP2015-45086, July 2015, provides

comparisons of KI computed using the original cladding model with SFT = 488°F with the KI computed using the NESC cladding residual stressesThe equivalent SFT for the NESC cladding stress can be

obtained using the FAVOR software to determine the SFT that matches NESC cladding stress KI

After the new version of FAVOR is released (and when funding is available), MRP & BWRVIP will resume work on the small flaw issues using a SFT appropriate for the circumferentially-oriented small flaw being analyzed in the “Appendix G small surface flaw” issue


Together…Shaping the Future of Electricity

annual nrc and industry materials technical exchange meeting … · 2016-08-10 · 2016 materials...

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