materials apc (mapc) - amazon s3...concept from bwrvip d. czufin, tva a. demma, epri a. mcgehee,...

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

Scot Greenlee, Exelon, Technical Chair

Kurt Edsinger, EPRI

Wednesday, August 31, 2016

Materials APC (MAPC)Afternoon Session Materials

Date: August 19, 2016

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

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


Anne Demma

Program Manager, EPRI

MAPC Meeting


Progress on Knowledge

Transfer / Training

Examples from MRP



General Approach

MRP identified some trainings of interest, prioritized, and moved

forward with developing 3 trainings in 2016-17

MRP will identify other top topics for future knowledge transfer

MRP members will vote on the topics and prioritize them

For each of the topics, we will determine the recommended

mechanism for the knowledge transfer

We 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


Current Examples from MRP

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

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

CBT developer attended workshops

Two EPRI Materials staff trained on e-learning and training software used by CBT developer (i.e., Storyline 2)


Together…Shaping the Future of Electricity


Andy McGehee

Senior Program Manager

MAPC Meeting


Technology Transfer

ConceptBWRVIP “Turbo Tax”



Current Way We Transfer Technology (Guidance)

Assessment Inspection Repair/Replace Mitigation

Component (I&E) Guidelines Guidelines Design Criteria Recommendations

Core 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 Update

Carl Larsen

Nuclear Asset Protection


Outline

• Recent important events

• Delivering the Nuclear Promise and engineering programs


Recent Important Events - PWRs

1. 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 - BWRs

1. 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, Entergy

Dan Wells, Program Manager, Chemistry

Materials Action Plan Committee Meeting


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 RS

Chemistry 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 Status

Review 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 ImplementationAfter 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 Metal

CS, SS, Ni-based alloys

coolant Fe2+X

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 (EPRI), PSCR Program Manager

Materials Degradation & Aging Action Plan Committee Meeting

Wednesday, August 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

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-13a

P-AS-13b

B-AS-09

B-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-02

P-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-47

P-AS-04, 05

P-DM-10

B-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-1MRFA-2

ARRM 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-12

P-AS-20

P-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-24

P-AS-30

P-AS-41

P-AS-42

P-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-02

B-AS-07

B-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-09

P-AS-40

B-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-46

P-DM-15

P-DM-16RFA-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 690

b) 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-01: IASCC of Internals, IMT gaps: PWR-AS-14a & b; BWR-AS-10, 11 & 26

• 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


• 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


• Will provide the results and participate in the development of EDF PWSCC Local model for plant

aging management

• Provide the mechanisticl 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/model

o Identify data gaps

o Perform IASCC CGR tests to fill the gaps (if necessary)

o Develop IASCC model of Ti-321SS for VVER


• 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 Deliverables

1. 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

• Program reorganization for better realignment

Collaboration, Coordination, & Integration


Member Engagement

Strong engagements with US and international members. We can and should

do more:

• 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)

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


PSCR Organizational Update

PSCR Leadership

TAC Chair: Jim Cirilli, Exelon

Technical Chair: Ian Armson, Rolls Royce

EPRI Program Manager: TBD

• TG Lian has been assigned to Europe

Update on program reorganization


Kevin Hacker, Dominion

NDE Chair

Steve Swilley, EPRI

NDE Program Manager

Materials APC


NDE Program

Quick 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, FENOC

WRTC Chair

Greg Frederick, EPRI

WRTC Program Manager

Materials APC


WRTC Program

Quick Summary



WRTC Program Summary - Outline

WRTC Strategic Plan

Key 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



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

International

Research 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

1

Research 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

1

Research 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



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



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 Mitigation

Steve 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 Activities

1. Weldability Solutions – Alloy screening

• Simplified field deployable DDC

screening test for high Cr alloys

• Standards for screening test

2. 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

Eff

ective

He

at In

pu

t (K

J/c

m)

Helium Concentration (appm)

JOG-GTAW-BM - No Cracking JOG-GTAW-BM - Cracking

JNES-GTAW-BM - No Cracking JNES-GTAW-BM - Cracking

JNES-GTAW-WM - No Cracking JNES-GTAW-WM - Cracking

LBW-BM - No Cracking LBW-BM - Cracking

LBW-WM - No Cracking LBW-WM - Cracking

LBW - Remelt Trials - No Cracking LBW - Remelt Trials - Cracking

JNES-GTAW-WM - GBD LBW-BM - GBD

LBW - 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 Attended

Fundamentals of Welding •OPG: Whitby, Ontario (Assisted by York Chan) September 2015, ~30

Operational Leakage Workshop •Palm Coast, FL

•New Orleans, LA

June 6, 2015 ~50

December 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 ~ 25

June 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 Date

Training for Repair/Replacement

Engineers (ASME Section XI)

•Duke, Charlotte, NC

•TVA, Chattanooga, TN

•Xcel Energy, Monticello - MN

April 19 - 22, 2016

August 1- 4, 2016

August 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 (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 Overview

2. Introduction to Welding, Brazing, and Fusing

3. Introduction to Weld Configurations, Considerations, and Defects

4. Introduction to Materials and Welding Metallurgy- Part 1

5. Introduction to Materials and Welding Metallurgy - Part 2

6. Introduction to Welding Codes

7. Introduction to ASME Section IX

8. Weld Filler Material

9. Welding Program Part 1

10.Welding Program Part 2

11.Alternate Repair Methods

12.Introduction to Inspection Techniques



Development 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

– Non

– Nuclear Promise

Risk informed Repairs, Case N-662-1

Identify best practices for repair and replacement activities toward reducing costs

– 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


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 much

Li 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 yearEach 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

Co

rro

sio

n R

ate

(m

g/d

m2)

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

UsageLi-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

“Plan A” (Reasonably Conservative)• Detailed project scopes developed for each technical item• 8 – 10 years• $8M - $10M

What is truly

necessary in the

face of no Li

availability?

“Plan B”

Phase 1: Qualification ahead of the PWR plant trial

Phase 2: PWR plant trial

Start of Phase 2


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 Palm

EPRI

MAPC Meeting


Environmentally Assisted

FatigueEPRI Overview



Content

Background

EPRI EAF Perspective

EPRI Approach to EAF

Current 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 Roadmap

Environmentally 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 Fatigue

Sponsorship & OrganizationEPRI 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 –

experience

Chair, International EAF committee coordinating EAF testing world wide. Committee spear heading –

testing of full, prototypical test fixture.

mailto:[email protected]




Analytical Committee Activities

Existing 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


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


• Guidelines for Addressing Environmental Effects in Fatigue Usage Calculations

• Issued December 2012




Efforts directed by EPRI EAF Focus Group comprised by industry fatigue practitioners

Two 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 Roadmap

Understand 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


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



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



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 transient

Theory: 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

Positive

strain 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 +0.0006

I1-2 325 0.60 Constant Slow-fast 600 480 1000 +0.001 -0.0025 +0.0006

I1-3 100 0.60 Constant Fast-fast 600 480 1000 -0.001 +0.0025 -0.0006

I1-4 325 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 +0.0006

NI2-2 100-325 0.60 In-phase Slow-fast 600 480 1000 +0.001 -0.0025 +0.0006

NI2-3 100-325 0.60 In-phase Fast-fast 600 480 1000 -0.001 +0.0025 -0.0006

NI2-4 100-325 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 Results

Test 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

Fati

gu

e L

ife i

n P

WR

en

vir

on

men

t

Prediction by NUREG/CR-6909 Rev.1

Prediction by JSME S NF1-2009

Experimental Fatigue life100ºC

Slow-Fast

325ºC

Fast-Fast325ºC

Slow-Fast

100ºC

Fast-Fast

Out of Phase

Slow-Fast

Out of Phase

Fast-Fast

In Phase

Slow-Fast

In Phase

Fast-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 hours

Regardless 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, 2016

Presentations made on analytical and experimental activities

Favorable feedback obtained

NRC 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