materials research update for lto - nrc: home pageb-as-32 assessment of core plate rim hold down...

Robin Dyle EPRI

NRC/Industry Meeting on LTO June 20, 2013

Materials Research Update for LTO

2 © 2013 Electric Power Research Institute, Inc. All rights reserved.

Outline

• NEI 03-08 History and Impact

• Use of MDM and IMT

• Materials Activities that address LTO


Materials Initiative (NEI 03-08) Expectations • The body of materials work conducted across the industry

will be forward-looking and coordinated, resulting in fewer unanticipated issues that could consume an inordinate level of resources and divert focus from an orderly approach to managing materials

• This initiative will enhance the issue programs’ ability to rapidly identify, react and effectively respond to emerging issues – OE is reported and industry responds

• Every utility will fully participate in the implementation of the materials management activities applicable to its plants

• INPO verifies implementation - excellence is target

• NRC interactions routine and frequent – no surprises


Issue Programs in Initiative

•EPRI –BWRVIP –MRP –SGMP –NDE –Primary Systems Corrosion Research –Water Chemistry Control

•PWROG Materials Subcommittee


BWR and PWR, Materials Program Focus


Progression of Environmental Cracking Damage

Development of Crack Nucleation Sites

Development of Crack Precursors

Development and Linkage

of Small Cracks

Growth of Large

Cracks

A large fraction of component life is spent in Stages I-III

|--------------Corrosion Research---------------| --Issue Programs--|

cost

----Detection---

Limit


Integrated Materials Issues Strategic Plan

• Provides Systematic Approach to Managing Materials Issues – Identify vulnerabilities – Assess condition (inspect & evaluate) – Mitigate degradation initiation and propagation

mechanism – Repair or replace as required

• Approach Used: – Degradation Matrix and Issue Management Tables

• Degradation Matrix and Issues Management Tables to be maintained as living documents


Industry Materials Degradation and Issue Management Table Approach

Develop a fundamental understanding of the degradation phenomena/mechanisms

Perform operability and safety assessments Develop Inspection and evaluation guideline Evaluate available mitigation options Develop repair & replace options Monitor and assess plant operation experience Obtain regulatory acceptance

Materials Degradation Matrix (MDM) and Issue Management Tables (IMT) are effective materials aging management tools in support of industry’s Materials Degradation and Issue Management Initiative

MD

M

IMTs


Materials Degradation Matrix (MDM) Revision 1

• MDM provides a comprehensive listing of potential degradation mechanisms for existing LWR primary system components

• Assesses the extent to which applicable degradation mechanisms are understood

• Evaluates the state of industry knowledge worldwide associated with mitigation of applicable degradation mechanisms

• Documents the results of an expert elicitation process • Proactively identifies potential challenges to avoid surprises


MDM Revision 1 Strategic Issues

• Environmental Effects on Fracture Resistance • Environmental Effects on Fatigue Life • SCC of Ni-Base Alloys • SCC of Stainless Steels • Effect of Fluence on SCC Susceptibility and SCC Crack

Growth Rates

NRC PMDA reached the same conclusions


2010 MDM Revisions

•2010 Revisions of the MDM (Rev. 2) address: – 80-year operations (Long-Term Operations or “LTO”) – Updates on identifying degradation mechanisms – Recent operating experience – Industry progress in addressing LWR materials issues – Most Gaps revised to keep contents up to date

• NRC and DOE Involved in meeting on revision • NRC EMDA 2012 Draft starts with MDM

– NRC has accepted MDM/IMT value


Identify Strategic Issues in Materials Degradation

• Major LTO issues: –Increased EOL Neutron Fluence

(RPV integrity, high fluence effects on austenitic SS, expanded regions of neutron effects)

– Increased Fatigue Cycles (with focus on environmental effects)

–Late-Life SCC Initiation and Stress Improvement Technique Stability

–Steam Generator Fouling / Corrosion / Long-Term Management

• Other major issues: –Effect of environment on fracture properties

–SCC initiation factors (cold work, welding effects, PWR system oxygen ingress)


MDM Results---- PWR Reactor Internals


MDM Results for PWR: – Neutron Irradiation Effects

Low Alloy Steels 2008 2010 LTO

Irradiation embrittlement Irradiation embrittlement Irradiation embrittlement

Austenitic Stainless Steels 2008 2010 LTO

–IASCC –Irradiation embrittlement –Void swelling



High Strength Stainless Steels and Alloy X-750 2008 2010 LTO

A-286 –IASCC –Irradiation embrittlement

–IASCC –Irradiation embrittlement

–IASCC –Irradiation embrittlement

X-750 –IASCC –Irradiation embrittlement


MDM Results---- BWR Reactor Internals


Issue Management Tables


Issue Management Table(s)

• Failure Consequences Evaluation Basis – Component failures considered in the IMT may be beyond current design or

licensing basis. All reasonably postulated events will be considered, solely to establish relative importance of postulated failures. No new design bases will be implied by the failures considered.

• An adverse consequence is failure of a component that results in one or more of the following:

• Inability to come to safe shutdown • Precludes maintenance of coolable core geometry • Loss of core cooling effectiveness • Loss of reactivity control • Reduction/elimination of critical Instrumentation availability • Causes a design basis accident • Significant economic impact • Significant on-site/off-site radiation release • Jeopardy to personnel safety • Breach of the reactor coolant system pressure boundary • Breach of the fuel cladding


Issue Management Table(s), cont’d

• Used to help focus efforts where most needed

• Used by Issue Programs set priorities and funding

• Used to inform NRC of key areas that require efforts

• Used to inform the NRC EPMDA effort

• Gives insights for coordination with DOE on research needs


Example IMT – PWR


Example Gap Description – PWR


PWR IMT Update

• Product issued as MRP-205, Rev 2 (1021024) • 76 Currently Open Gaps (total same as for Rev 1)

– 16 new R&D gaps identified • Majority tied to consideration of longer service life (>60 years)

• 1 High Priority, 5 Medium Priority, and 10 Low Priority – High Priority = P-I&E-22 – “Appendix VIII Compliance”

– 16 previous R&D gaps closed – 30 High Priority Items (7 elevated from Medium plus 1 new)


BWR IMT Update

• Product issued as BWRVIP-167NP, Rev 2 (1020995) • 45 Currently Open Gaps (3 less than for Rev 1)

– 10 new R&D gaps identified • Majority tied to consideration of longer service life (>60 years)

• 3 High Priority, 4 Medium Priority, and 3 Low Priority – High Priority = B-AS-29 “Steam Dryer Evaluation

Methodology” – 13 previous R&D gaps closed

• 18 High Priority Items (4 elevated from Medium plus 3 new)


PWR LTO Summary

• Neutron Fluence Effects – RPV embrittlement – SS materials data for >60 years

• Threshold stress • Reduction in toughness • Void swelling

– Impact on core periphery materials • Fatigue Usage • Steam Generator Corrosion Limits

– FAC impact – Number of cleaning cycles


BWR LTO Summary

• Neutron Fluence Effects – RPV embrittlement – Irradiation effects on LAS resistance to environmentally

assisted cracking – SS materials data for >60 years – Impact on CASS reactor internals – Impact on nickel alloys – Irradiated material welding

• Fatigue Usage • Late-life SCC Initiation

– Impact of oxide formation/environment exposure


Publication of EPRI MDM and IMTs

• EPRI MDM Rev-3 (May 2013), EPRI Report # 3002000628

• BWR and PWR IMTs were updated to reflect MDM extension to 80 years •BWR-167NP, Rev-2, EPRI Report # 1020995 •MRP-205, Rev-2, EPRI Report # 1021024

• BWR and PWR IMTs revisions to address Rev 3 of MDM are in review

• Available at www.epri.com

http://www.epri.com/�


BWRVIP

Materials and Internals


BWRVIP Guidelines to Manage Degradation Assessment Inspection Repair/Replace Mitigation

Component (I&E) Guidelines Guidelines Design Criteria Recommendations

Core shroud BWRVIP-76, R1 BWRVIP-03 BWRVIP-02-A/-04-A BWRVIP-62, R1/-190

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

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

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

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

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

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

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

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

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

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

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

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

Access hole cover BWRVIP-180 BWRVIP-03 TBD BWRVIP-62-, R1-190

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


Current IMT High Priority Gaps

Gap 2012BWRVIP Tasks2013

BWRVIP TasksB-AS-18 Jet Pump Degradation Management 2.17 2.17B-AS-10 Assess Impact of High Fluence SCC Crack Growth Rates 2.5, 2.6, 2.13, 2.34 2.5, 2.6, 2.13, 2.34B-MT-02 ECP Measurement, Estimation, and Demonstration B-MT-05 Water Chemistry Optimization for Power Operation, Startup and Shutdown

3.1, 3.10 3.1, 3.10

B-AS-26 Assess Alloy X-750 SCC Susceptibility 2.15 2.15B-I&E-03 NDE Capability: Shroud & Shroud Support Locations 2.27 2.27B-AS-29 Steam Dryer Evaluation Methodology 2.8 2.8B-DM-06 Environmental Effects on Fracture Resistance PSCR PSCRB-RR-02 Develop Welding Processes for Repair of Irradiated Material 2.21 2.21B-RG-09 Management of License Renewal Issues Many ManyB-AS-32 Assessment of Core Plate Rim Hold Down Bolts 2.18 2.18B-RG-05 Evaluation of Remote EVT-1 2.27, 2.28 2.27, 2.28B-MT-04 On-Line NMCA Deposition Effectiveness and Implementation

3.1, 3.5, 3.93.1, 3.5, 3.8, 3.14, 3.17, 3.18

B-AS-07 Environmental Fatigue Issues: Pressure Boundary 2.31 2.11, 2.31B-AS-09 Assess Impact of High Fluence on Fracture Toughness 2.5, 2.7 2.5, 2.7


Current IMT Medium Priority Gaps


BWRVIP TasksB-AS-27 Alloy 182/Creviced Alloy 600 SCC SusceptibilityB-AS-28 Impact of BWR Nozzle Penetrations on Pressure-Temperature Limit Curves

BWROG BWROG

B-MT-01 Development of Alternative Mitigation Technologies MRP (peening) MRP (peening)B-I&E-02 NDE Capability: Hidden Locations (Thermal Sleeves & Piping) 2.27 2.27B-AS-05 Fluence Spectra Effects: C&LAS DOEB-I&E-01 NDE Capability: Core Plate Rim Hold Down BoltsB-AS-15 High Cycle Fatigue: Reactor Internals 2.31 2.31B-RR-05 Develop Alternate High-Strength Materials PSCR (ARRM)B-RR-08 Availability of Laser Welding for Repairs to Highliy Irradiated Components

2.21 2.21

B-DM-07 Chloride Transient Effects on Low Alloy Steel Crack Growth Rates 3.10 3.10

B-AS-30 Material Surveillance Program Implementation for 80 Year Service LivesB-RG-08 Reactor Pressure Vessel Material Surveillance Program Implementation for 80 Year Service LivesB-AS-33 Equivalent Margins Analysis for BWR NozzlesB-AS-22 High Cycle Thermal Fatigue - Piping LocationsB-AS-11 Assess Fast Reactor Data Applicability to BWR Environment 2.5 2.5


Current IMT Low Priority Gaps


BWRVIP TasksB-AS-31 BWRVIP-47-A (CRGT) Re-Inspection RequirementsB-AS-20 Assess Non-Safety LocationsB-AS-14 Environmental Fatigue Effects: Reactor Internals 2.31 2.31B-I&E-08 Develop Inspection and Evaluation Guidance for RepairsB-I&E-06 NDE Capability: CASS Weldments NDEC NDECB-AS-12 Thermal & Irradiation Embrittlement: Synergistic Effects (on CASS BWR Reactor Internals)B-DM-03 Low Temperature Crack Propagation PSCR PSCRB-I&E-09 Examination Technologies for Detection of Stress Relaxation in Reactor Internals ComponentsB-DM-10 Long Term Stress Stability (LTO Based Gap)B-DM-08 Long Term Neutron Fluence Effect on Low Alloy Steel Cracking SusceptibilityB-DM-09 Long Term SCC Susceptibility (Late Life Initiation)B-RG-10 R.G. 1.161 and ASME Section XI Appendix K Stress Intensity Factor Equation Non-conservatisms


MRP - PWR

Materials and Internals


MRP Objectives

• Develop fundamental understanding of degradation mechanisms and conduct generic operability and safety assessments

• Develop Inspection and Evaluation guidelines

• Evaluate mitigation, and repair/replace options

• Monitor, evaluate and feedback worldwide Operating Experience

• Develop technical bases for regulatory review and to support licensees

• Leverage industry resources


Key 2012-13 Deliverables

• In 2012 • PWR Reactor Internals Inspection Standard (MRP-228, Rev. 1) • Assessment of Residual Heat Removal Mixing Tee Thermal Fatigue

in PWR Plants (MRP-192, Rev. 2) • Boric Acid Corrosion Guidebook, Revision 2: Managing Boric Acid

Corrosion Issues at PWR Power Stations (MRP-058, Rev. 2) • Recommended Analytical Solutions and Approaches for Alternate

PTS Rule (10CFR50.61a) Implementation • In 2013

• Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-335, Rev. 1)

• Resistance of Alloys 690, 152 and 52 to Primary Water Stress Corrosion Cracking (MRP-237, Rev. 2)

• Probabilistic Fracture Mechanics (PFM) Analysis of Reactor Vessel Shell Forgings with Quasi-Laminar Indications

• PWR Issue Management Tables report (MRP-205, Rev. 3)


MRP Guidelines Status

Guideline Title TAC NEI 03-08 MRP Rpt #

Current Rev #

G/L Report

#

Last Pub. Date

Next G/L

Review Date

Generic Guidance for Alloy 600 Management

AST Man. 126 0 1009561 November 2004 -

Primary System Piping Butt Welds Inspection and Evaluation Guidelines

AST Man. 139 1 1015009 December 2008 -

MRP-139, Revision 1 Interim Guidance on Rescission of MRP-139, R1 Mandatory Requirements with Implementation of Code Case N-770

AST Man. Letter 2010-046

0 - January 2011 -

Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines

TS Need

146 1 1022564 June 2011 -

Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines – Supplemental Guidance

TS Need

146S 0 1018330 January 2009 2013

Assessment of RHR Mixing Tee Thermal Fatigue in PWR Plants

TS GP 192 2 1024994 September 2012 2014

Materials Reliability Program: PWR Internals Inspection and Evaluation Guidelines

AST Man. 227-A 0 1022863 December 2011 -

Materials Reliability Program: Inspection Standard for PWR Internals

Insp Need 228 1 1025147 December 2012 2014

Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP) Guidelines

TS Need

326 0 1022871 December 2011 2013


2013 Key Research Areas

• Extremely Low Probability of Rupture (xLPR) is the largest project in MRP • Address PWSCC in piping previously approved for leak-before-break

• Reactor Internals Aging Management, Inspection and Testing • Irradiated materials testing of internals; Zorita, Gondole, Halden • Support utility’s implementation of the current guidance • Incorporate lessons learned from initial component inspections

• Replacement Materials: Alloy 690 and its weld metals • Investigate technical basis of inspection regime for replacement components • Quantify and maximize PWSCC resistance of Alloy 690 components

• RPV Integrity • Ensure that appropriate analytical tools & correlations are developed to

analyze & model vessel integrity for safe & efficient operation through 80 years • Environmentally Assisted Fatigue

• Resolve differences between fatigue test results and operating plant history


Project Summaries


Irradiated Materials Testing

• EPRI Irradiated Materials Testing and Degradation Models Roadmap

• Collaborative Research Programs – Zorita Internals Research Project

• Additional Zorita Materials Projects – Halden Research Program – IASCC Data Compilation and Analysis

• Primary Systems Corrosion Research Projects • MRP Projects • BWRVIP Projects


EPRI Irradiated Materials Testing and Degradation Models Roadmap


Zorita Internals Research Project

MRP, BWRVIP, and PSCR


Zorita Internals Research Project (ZIRP)

• Increased understanding of irradiation effects on: • Mechanical properties: tensile strength, fracture toughness,

crack initiation and growth • Microscopic properties: grain boundary chemistry and size,

void formation, and hydrogen and helium production • Project uses Zorita baffle plate material

Objective

• P-AS-14 (high): Fluence Impact on SCC of Stainless Steels (IASCC)

• P-AS-15 (medium): Void Swelling of Stainless Steels • P-AS-38 (medium): Fluence Impact of Stainless Steel

Mechanical Properties (Fracture Toughness and Tensile Strength)

IMT Gaps


Zorita Internals Research Project Current Status

• Radiation analysis complete for all 29 cycles • Temperature analysis for 8 to 9 representative cycles to be

completed shortly • Segmentation and sample cutting underway • Cask arrived at Zorita from Sweden in early May


Zorita Internals Research Project Cutting Plan – Baffle Plates

= 52 dpa= 32 dpa= 15 dpa= 11 dpa

41.2

2”

7.8”

6.67”

Plate A (41.22” wide)

Plate B (7.8” wide)

Plate C (7.8” wide)

Type 304 •Doses ranging from a few dpa to ~58 dpa •Thickness 28.6 mm •ZIRP uses Pieces B1, B2, and B3


Zorita Internals Research Project Cutting Plan – Core Barrel

Type 304 •Core barrel weld has been located •Weld material to be used by MRP, BWRVIP, U.S. NRC


MRP Projects Using Additional Zorita Materials

• Objective: Determine the combined effects of irradiation and exposure to elevated temperature on embrittlement of stainless steel welds and characterization of environmental effect on fracture toughness in irradiated stainless steel welds

• Project uses Zorita core barrel weld material • P-AS-13 (high): Thermal & Irradiation Embrittlement Synergistic Effects on CASS and

Stainless Steel Welds

Thermal and Irradiation Embrittlement and Environmental Effects Testing of Stainless Steel Welds

• Objective: Develop IASCC CGR data in irradiated stainless steel weld and HAZ materials for comparison to existing data for base materials

• Project uses Zorita core barrel weld material • P-AS-14 (high): Fluence Impact on SCC of Stainless Steels (IASCC)

CGR Testing of Irradiated SS Weld and HAZ Materials

• Objective: Evaluate IASCC crack initiation and crack growth rates and degree of void swelling in highly-irradiated (near end-of-life conditions) stainless steel base metal and welds

• Project uses Zorita baffle plate & core barrel weld material • P-AS-14 (high): Fluence Impact on SCC of Stainless Steels (IASCC) • P-AS-15 (medium): Void Swelling of Stainless Steels

Determination of IASCC CGR, Initiation Rate, and Void Swelling in Zorita Material after Post-Reactor Irradiation


BWRVIP and PSCR Projects Using Additional Zorita Materials

• Objective: Develop data to help establish the SCC growth rate disposition curves and understand the relationship between fracture toughness (FT) and dose for austenitic stainless steels in the BWR environment

• Project uses Zorita core barrel base, weld, and HAZ materials • B-AS-09 (high): Assess the Impact of High Fluence on Fracture

Toughness • B-DM-06 (medium): Environmental Effects on Fracture Resistance

Crack Growth Rate & Fracture Toughness Testing of Weld Materials (BWRVIP)

• Objective: Determine the applicability of using data generated from smaller specimens to assess the performance of thicker internals components

• Project uses Zorita baffle plate material

K-size Effect Testing (PSCR)


Halden Research Project

MRP, BWRVIP, FRP, and PSCR


Halden Research Program Plant Aging and Degradation •EPRI participates in the Halden program which is supported by members from 18 countries and more than 100 organizations

•Halden Material Testing Program – Investigations are performed in representative thermal

hydraulic, coolant chemistry, and radiation conditions – On-line instrumentation and control of

• Coolant temperature • Water chemistry (H2 / O2) • Specimen stress level • Constant load vs cycling

– Samples manufactured from materials retrieved from power reactors


Halden Research Program 2012 to 2014 Test Plan Highlights

Task Objective Applicability

2.1.a. BWR Crack Growth Rate Study

•Generate long-term CGR data in irradiated specimens in BWR conditions •Determine the benefits of HWC in mitigating cracking in irradiated materials with different fluence levels

BWR

2.1.b. PWR Crack Growth Rate Study

•Generate long-term CGR data in irradiated specimens in PWR conditions •Determine the effects of hydrogen concentrations, Li/B ratios, and/or Zn additions and the benefits of Post-Irradiation Annealing (PIA)

PWR

2.1.c. Irradiation of CW 316 SS CTs for Future PWR Tests

•Irradiate CW 316 SS CT specimens to higher (4 to 5 dpa) dose for future PWR CGR tests

PWR

2.1.d. Crack Initiation (Integrated Time-to-Failure) Study

•Evaluate the benefits of HWC in mitigating the initiation of cracks irradiated (12 dpa) 304L SS tensile specimens •Test under BWR HWC conditions and PWR conditions (including with high Li (3.5 ppm) and with Zn additions)

PWR/BWR

2.1.e. Irradiated Materials Characterization

•Obtain detailed information on the irradiation history and mechanical and microstructural properties of materials being used in IASCC studies

PWR/BWR

2.2 Stress Relaxation Study •Establish the technical feasibility of on-line irradiation stress relaxation measurement during irradiation •Measure the irradiation stress relaxation of materials used in LWRs and benchmark the irradiation stress relaxation data to irradiation creep tests

PWR/BWR

2.3 RPV Integrity Study •Establish proper correlations between the neutron embrittlement data from sub-size specimens with those of standard size

PWR/BWR


Halden Research Program IASCC: Crack Growth Program

Objectives

• Measure on-line cracking response of materials retrieved from commercial reactor components

• Compare & quantify CGRs in BWR (280-290 °C, O2 and H2) vs. PWR (320-340°C, Li, B, H2) conditions

• Study cracking response as affected by – Stress intensity (K) level – Fluence – Microstructure – Mechanical properties (yield strength) – Flux


Halden Research Program IASCC: Time-to-Failure program

Objective • Determine effectiveness of HWC

in reducing susceptibility to the initiation of cracks in irradiated material

Experimental • Miniature tensile specimens

prepared from 304L SS • Load (up to >100% of YS)

applied by means of bellows • On-line monitoring of specimen

failures (LVDTs) • Compare number of failures

under HWC & NWC conditions • Study in PWR environment is

planned


IASCC Data Compilation and Analysis

PSCR, BWRVIP, and MRP


EPRI IASCC Data Compilation & Analysis Objectives and Scope

Recommend crack growth models and disposition curves Crack growth models and disposition curves for

irradiated stainless steels in BWR and PWR environments

Final report will be issued after review by PSCR, MRP, and BWRVIP

Convene an Expert Panel

Review, screen and categorize the available data using consensus screening criteria

Panel includes principle investigators, selected industry experts, and vendors.

Compile crack growth rate data on irradiated stainless steels CIR-fast reactor irradiated materials

Halden-LWR irradiated materials & in-reactor tests BWRVIP and MRP irradiated materials

Literature data (ANL NUREG reports, JNES reports)


EPRI IASCC Data Compilation & Analysis Expert Panel Status

• Database includes more than 1600 test segments including those under cyclic loading

• For IASCC crack growth rates only test segments under constant load or periodic partial unloading were considered

• Test segments ranked on a scale of 1 (best) to 5 (worst) after examining the raw data from crack length vs. time plots

• Only data ranked from 1 to 3 was considered suitable for development for models for BWR NWC, HWC and PWR environments


Low-ECP Residuals vs. Average Rank

-3

-2

-1

0

1

2

3

0 1 2 3 4 5 6Average Rank

log(

Mod

el C

GR

) - lo

g(C

GR

310

°C)

Fitted Low-ECP Data, Rank < 3

Data Not Fitted, Avg. Rank ≥ 3


EPRI IASCC Data Compilation & Analysis Next Steps

• Work with the EPRI Expert Panel to produce and document the final consensus PWR and BWR IASCC models from the extensive EPRI IASCC database that was compiled and ranked by the Expert Panel in earlier phases of this Project

• A revised version of the draft report on the IASCC database and both low ECP (PWR and HWC) and high ECP (NWC) IASCC models, incorporating all revisions requested by the Expert Panel and EPRI reviewers will be prepared by November 2013

• The draft report will be sent for PSCR, MRP and BWRVIP review in December 2013

• Final report will be published in the second quarter of 2014 after comment resolution

• The report will provide the technical basis for crack growth disposition curves for irradiated BWR and PWR stainless steel internals


PSCR: Mechanistic understanding on IASCC

• Identify the key factors influencing IASCC initiation and crack growth.

• Understand the linkage between irradiated microstructures and IASCC

• Confirm the processes that lead to occurrence of IASCC

Objectives

• Lack of understanding on IASCC • No mitigation strategies available • Uncertainty on reliability of components for LTO

Gaps


Closely collaborating with US DOE, to perform fundamental researches for better understanding of IASCC mechanisms Identification of Key Factors Affecting IASCC of

Austenitic Alloys in LWR Core Materials Establishing a Cause-and-Effect Relationship

between Localized Deformation and IASCC APT and Nano-SIMs Characterization of

Proton- and LWR Neutron Irradiated Stainless Steels

Investigation of Post Irradiation Annealing as a potential strategy for mitigating IASCC of reactor core structural components

PSCR: Mechanistic Studies on IASCC

Nishioka et al. J. Nucl. Sci. Techn. 45 (2008) 274.


Assess the role of solute additions, in particular the roles of C, Mo, Ti, Nb, Cr+Ni and P, on crack growth (CGR) and crack initiation (CI).

Understand the linkage between irradiated microstructures and CGR/CI for solute addition and commercial alloys, also effects of CW and dose.

Determine the predictive capability of crack initiation due to proton irradiation, compared to crack initiation due to neutron irradiation.

Investigate the role of localized deformation on the IASCC susceptibility in neutron irradiated materials.

PSCR: Mechanistic Studies on IASCC Identification of Key Factors Affecting IASCC

0

10

20

30

40

50

60

0

0.2

0.4

0.6

0.8

1

1.2

1.4

5.5 (AS13) 10.2 (AS17) 47.5 (AS22)

RA

(%) o

r %IG

Elon

gatio

n (%

)

Dose, dpa (Sample)

BWR NWC

NA

IASCC initiation crack growth


PSCR: Mechanistic Studies on IASCC IASCC Initiation Mechanism Study

Establish direct evidence of a cause-and-effect relationship Site-specific approach to confirm whether an IASCC crack can be

initiated at a pre-characterized site Direct measurement approach to determine the degree of localized

deformation at the crack sites

Measurement of displacements due to plastic and elastic strains • Digital image correlation measurement for in-plane displacement • Confocal microscopy for out-of-plane displacements • EBSD for elastic stresses

Development of IASCC mitigation strategy Focus on removing or reducing the degree of localized deformation

– Post-irradiation annealing to remove the dislocation loops – Cold-work to reduce the degree of localized deformation – Precipitation-strengthened alloys to prevent localized deformation


PSCR: Development of Advanced Radiation Resistant Materials (ARRM)

• Develop the next generation of materials for in-core structural components and fasteners.

• Determine the degradation-resistance of the current commercial alloys

• Determine the degradation-resistance of the new advanced alloys

Objectives

• No resistance materials for replacement of RI components

• No resistance materials for new plants (RI components)

Gaps


PSCR: Development of Advanced Radiation Resistant Materials Program (ARRM)

• Existing alloys used for LWR reactor internals have been known to degrade in less than 40 years

• Component replacement may be required in some reactors

• Advanced materials will allow reactors to operate with greater efficiency


PSCR: Development of ARRM

ARRM Project EPRI and the U.S. Department of Energy (DOE) are initiating a global,

collaborative research effort to develop the next generation of materials for in-core structural components and fasteners.

The two primary research goals are: – By 2022, to develop and test a degradation-resistant alloy that is

within current commercial alloy specifications – By 2024, to develop and test a new advanced alloy with superior

degradation resistance 10-year project, estimated $12 -15 M work-scope Project work-scope defined by sponsors and managed by EPRI as was

done in the CIR and NFIR programs


PSCR: Development of ARRMs Future Steps • EPRI and DOE have published “Critical Issues Report

and Roadmap for Advanced Radiation Resistant Materials Program”, (EPRI Report 1026482, December 2012)

• Both organizations will initiate a long term (~ 10 years) collaborative research program to develop and qualify more radiation resistant materials based this report

• We are looking for partners (vendors, research organizations, utilities, regulators) to join us in this important effort

• Organizations of the program will be similar to other EPRI collaborative programs (CIR and NFIR)


MRP: Gondole Void Swelling Irradiation and Testing

• Increase accumulated dose to ~30 dpa for virgin samples • Provoke swelling on other materials to determine kinetics of swelling • Investigate possible existence of threshold temperature for swelling

Objective (Phase 2)

• P-AS-15 (medium): Void Swelling of Stainless Steels

IMT Gap

• Virgin PWR internals materials & pre-irradiated materials exposed for 5 cycles of irradiations • Virgin materials accumulated ~14 dpa • Pre-irradiated materials had up to 85 dpa

• Type 316: No significant swelling; several cases of densification • Type 304: 6 cases of significant swelling • Type 347 and NMF 18 CW: Significant swelling on each

• *Significant swelling/densification defined as change in volume > 5%

Phase 1 Results


MRP: Gondole Void Swelling Project Comparison to Void Swelling Model

•Cluster Dynamics model developed to characterize void swelling in austenitic stainless steels

•Model being benchmarked with Gondole data

Blue line segment: Modeling of in-service irradiation Red line segment: Modeling of test-reactor irradiation (Osiris, 360° C) Black squares: Gondole data (irradiation Phases 1 to 5)

__1E-7 dpa/s, 287°C, 20 appm He/dpa

__ 3E-7 dpa/s, 360°C, 10 appm He/dpa

Gondole Data

__1E-7 dpa/s, 330°C, 20 appm He/dpa

__ 3E-7 dpa/s, 360°C, 10 appm He/dpa

Gondole Data


MRP: Lithium Effects on IASCC Initiation

• Determine the effect of lithium (Li) on the rate of IASCC initiation for comparison to recent data generated by EDF suggesting increased Li concentration may enhance IASCC initiation rate

Objective


IMT Gap

• Material: Flux thimble tubes from Ringhals Unit (60 and 100 dpa) • Lithium levels 2.0 and 8.0 ppm • UCL and o-ring specimens

Experimental Design


MRP: Effect of Lithium on SCC Initiation Background

0

100

200

300

400

500

600

700

800

900

1 10 100 1000 10000 Time to failure (h)

Stre

ss (M

Pa)

C7 800 MPa, PWR, 340°C C4 700 MPa, PWR, 340°C C6 700 MPa,PWR, 340°C (fast loading)

C8 400 MPa, PWR, 340°C : not broken C3 500 MPa, PWR,340°C C9 700 MPa, PWR, 290°C

C2 600 MPa, PWR, 340°C C1 500 MPa, PWR high Li, 340°C Behavior Chooz A - 30 dpa ?

C12 400 MPa, PWR high Li, 340°C C8 350 MPa, PWR high Li, 340°C : not broken Behavior Chooz A - 30 dpa - High Li ?

No cracking

No cracking (SEM to be performed) No cracking

340°C 290°C

[Li]=3.5 ppm [Li]=2.2 ppm


MRP: Dynamic Strain Effects on IASCC Initiation Rates

• Compare existing results from static-loaded tests to tests conducted using dynamic loads representative of PWR transients to better understand EDF baffle bolt experience and IASCC test observations

Objective


IMT Gap

• Project hosted by NUGENIA; led by EdF • 13 organizations collaborating to provide materials, in-pile & out-

of-pile testing, modeling

Participation


MRP: Dynamic Strain Effects on IASCC Initiation Rates Background Few data on the effects of temperature or stress variation on IASCC sensitivity Halden Project showed an association of specimen failures with test interruptions and specimen re-loading. Data obtained on non-irradiated materials show an effect of dynamic loading on initiation Possible role of dynamic straining in IASCC initiation.

By Torril Karlsen - Halden


MRP: Dynamic Strain Effects on IASCC Initiation Rates Experimental Approach

1. Numerical analysis of the most severe transients for PWR reactors

2. Post-irradiation mechanical testing of stainless steel materials (out-of-pile tests) • SA 304 and CW 316 irradiated in experimental or commercial reactors • 4 and 20 dpa (typical range for IASCC failures) • PWR environment using transients identified in Phase 1

3. In-flux testing of pre-irradiated materials in a material test reactor • Neutron flux with load increments or temperature increment to simulate the shut-

down and start-up of PWR • Tests will also be performed without load or temperature increment for comparison

4. Tests on non-irradiated materials • Assess tapered specimens with representative CW • Sensitivity to transients on SCC initiation

5. Modeling of the effects of transients at grain scale


BWRVIP : Testing of Unirradiated and Irradiated X-750 and XM-19 Materials

• Develop additional mechanical property data, SCC susceptibility, and CGR data to determine the long-term viability of currently installed materials.

• Irradiation of materials to be performed at INL-ATR NSUF

Objective

• B-AS-26 (high) High-Strength Alloys • B-RR-05 (medium) Alternative High-Strength

Materials

IMT Gaps


Background

• Alloys X-750 and XM-19 are used in many structural applications in BWRs ranging from original equipment to modifications and repair hardware – SCC and fracture toughness data in

BWR water chemistry conditions are rather limited, particularly when exposed to neutron irradiation

• A multi-year program has been developed to examine the SCC and fracture toughness behavior of these materials under a variety of BWR conditions


X-750 and XM-19 Data Needs

• SCC growth rates in NWC and HWC

• J-R fracture resistance in air and high temperature water

• Crack initiation • Effects of neutron fluence • Effects of microstructure

variability


Project Structure

• GE Global Research – Material characterization

• Micro-macro mapping and SEM analysis – Unirradiated testing

• Tensile, crack growth and J-R tests • Crack initiation in constant load Keno test

• Idaho National Laboratory – Advanced Test Reactor – Confirmatory unirradiated testing (for comparison to GE-GR) – Irradiation at three fluences – Post irradiation examination and testing

• TEM examination, tensile, crack growth and J-R tests


SCC CGR Testing - Purpose and Experimental Strategy • Evaluate different microstructures of Alloy X-750 and

compare to prior data • Evaluate effects of K, dK/da, corrosion potential & water purity • DCPD crack monitoring with anticipatory correction • BWR water: 2 ppm O2 or 63 ppb H2; 0, and 10 to 30 ppb SO4 • Corrosion potentials measured with ZrO2 / Cu2O & Pt electrodes • CT specimens tested in S-L, L-T and T-L orientation from 10 – 45 ksi√in

– GE : 0.5T CT – INL: 0.4T CT (size to match ATR capsule specimens)


Comparison of GE and INL CGR Tests of Alloy X-750


Preliminary Conclusions

• SCC CGR tests on Alloy X-750 – < 2X difference among all microstructures or orientations

• No evidence of accelerated CGR due to variations in microstructure

– Strong effect of corrosion potential (HWC) on reductions in CGR – INL confirmatory testing compares well with GE-GR

• BWRVIP-262NP: BWR Vessel and Internals Project, Baseline Fracture Toughness and Crack Growth Rate Testing of Alloys X-750 and XM-19 (Idaho National Laboratory Phase I). EPRI, Palo Alto, CA: 2012. 1025135.

• Fracture toughness tests on Alloy X-750 – Scatter in JIc and J-∆a is small – All specimens exhibit significant ductility


Alloy 690 and Weld Metals Testing


Background – Alloy 690

• In 2006, Bettis identified high CGRs (~10XE-07mm/s) for cold worked Alloy 690; about the same as for Alloy 600 (MRP-55)

•Assembled an Expert Panel and an International collaboration to

guide and conduct complementary and confirmatory research

•Monitoring of research by expert panel

•Use of well-established test procedures and characterized materials; many using materials from a central inventory

•Members include EPRI, NRC RES, Bettis/KAPL, EDF/MAI, UNESA, KAERI, MHI, Tohoku Univ. and many others

•Reviewed state of knowledge in 2008 (MRP-237, Rev.1); agreed to eleven knowledge gaps (next slide); an update of research published in 2013 (MRP-237, Rev. 2))


Alloy 690 PWSCC Knowledge Gaps

1. PWSCC susceptibility of HAZ (H) 2. Effect of weld defects in 52(M)/152 on PWSCC susceptibility (H) 3. Effect of weld composition & welding procedure (including dilution effects) on

PWSCC and LTCP (H) 4. Welding fabrication and repair effects on defect population, residual stresses

and susceptibility (H) 5. Reduced resistance to PWSCC due to thermo-mechanical processing of A690

(e.g. 1-D rolling) (M-H) 6. Resolution of contradictory CGR findings among labs for 52(M)/152 (M-H) 7. Relevance of thermo-processing modes (e.g. 1-D rolling) to plant installations

(M-H) 8. Importance of LTCP to operating plants for A690 and welds (M) 9. CGR flaw disposition curves for A690/52/152 (M) 10. Detailed information on actual replacement components in the field (L-M) 11. Crack initiation data on heterogeneously deformed A690 that has shown high

CGR Values(L-M)


EPRI Alloy 690/52/152 PWSCC Program Issue Statement

• Alloys 690/52/152 very resistant to PWSCC; however, PWSCC-related

issues remain: – PWSCC vulnerability for material with abnormal microstructure,

specific product forms, and to thermo-mechanical processing that could be present in the HAZ

– A52 fabrication defects could become a factor in initiation or growth of PWSCC cracks in welds or weld overlays.

– Guidelines for material procurement specifications and weld quality are needed to minimize concerns for PWSCC

• Resolving these issues can inform regulatory consideration of inspection optimization for replacement components such as the reactor pressure vessel head with Alloy 690 penetrations.


Alloy 690/52/152 PWSCC EPRI Program Plan (1/2)

• Alloy 690 PWSCC Degradation Characterization • Alloy 690 HAZ PWSCC Degradation Characterization • Alloy 690/52/152 Macro and Microstructural Mapping and

Strain Analysis • Alloy 690/52/152 Expert Panel • Guidelines for Alloy 690 Material Procurement • Technical Bases for Regulatory Relief for Alloys 690/52/152 • CGR Data to Develop Flaw Disposition Curves for Alloys

690/52/152 • Weld Metals PWSCC Degradation Characterization • PWSCC Resistance of Evolutionary A52 Compositions

Above tasks executed per roadmap schedule

shown on next slide


Alloy 690/52/152 PWSCC EPRI Program Plan (2/2)


CGR Testing

• A690 Testing Parameters – Product form – Microstructure – Heat treatment – Cold work and direction – Crack Orientation – Heat Affected Zone (HAZ); Dilution Zone – Temperature, Stress intensity, Environment

• Alloy 52/152 Testing Parameters – Weld configuration – Crack orientation – Effect of pre-existing cracks – Weld overlay (Alloy 52/52M over Alloy 182/82) – Temperature, Stress intensity, Environment


Summary Alloy 690 CGR Testing

• Relatively high CGRs measured in certain specimen orientations after appreciable (>20%) levels of cold work, CGRs reduce at lower cold work levels

• Clear evidence of effect of cold versus hot forging on CGRs • The role of banding in measured high CGRs has not been confirmed;

but effect is expected to be small • Tests to date do not show high CGR susceptibility in the weld HAZ, but

testing of such specimens is still limited • Weld and HAZ strains even after weld repair appear in the 10 to 15%

range (not a major concern), work continuing • Concern for dilution zone at interface with low-chrome material needs

to be addressed


Alloys 52/152 CGR Testing Current Status

• To date, only one laboratory has shown PWSCC CGRs of real concern for Alloy 152 (~5xE-08mm/s) – Substantial other data show CGRs in the 5xE-09mm/s

range • Some evidence is emerging that pre-existing weld flaws in

the form of hot cracks affect neither the propensity to PWSCC initiation nor to crack growth; more work is continuing


PWSCC Initiation Testing for Alloy 690 and Weld Metals • Initiation of PWSCC cracks and their growth

to detectable levels represents >80% of component life

• Current approach on inspection intervals based on growth of a detectable crack, no recognition for time for crack to initiate

• The NRC/MRP are now developing a probabilistic regulatory framework (xLPR) that will allow recognition of crack initiation times allowing optimization of inspection strategy

• Need PWSCC initiation test data to benefit from xLPR

Perform PWSCC initiation time testing for Alloy 690 and weld metals accounting for variables such as cold work, HAZ, surface conditions, weld design, heat-to-heat and vendor variations, and environment.

Develop PWSCC initiation test data to establish basis for alternate inspection intervals for Alloy 690 components in the xLPR framework.

Objective & Scope

Issue

Crack Nucleation

Crack Precursor

Short Crack

Crack Growth

Detectable Crack


PWSCC Initiation Testing Projects

• Advanced Nuclear Technology project on PWSCC initiation testing of Alloy 52 (2013-2016)

• EPRI funded project at KHNP for PWSCC initiation testing of Alloy 690 at 400ºC steam (2013-2016)

• MRP funded project on PWSCC initiation testing of Alloy 690 under primary water conditions (2014-2017)

• EDF/MAI project on PWSCC initiation of steam generator plugs

These will feed into the xPLR platform


Reactor Vessel


Motivations to Generate LWR Surveillance Data

• There is a paucity of PWR high fluence data above 5E+19 n/cm2 upon which to base an ETC for PWR operations to 80 years

• MRP Actions to Address Paucity of High-Fluence PWR Surveillance Data

1. Test CLB capsules at a higher fluence (CRVSP)

2. Design, fabricate, irradiate and test capsule(s) (PSSP)

• Allows selection of optimum materials to attain goals

• Requires larger initial capital outlay, but highly information-efficient


Coordinated Reactor Vessel Surveillance Program (CRVSP) • Coordinated PWR Reactor Vessel Surveillance Program

(MRP-326) – Produces high fluence data (3-10 x1019 n/cm2) from remaining

Appendix H capsules by 2025 to support future ETC development • CRVSP defers capsules only if can attain >3E+19 n/cm2 by 2025

– Implemented as a “Needed” guidance under NEI 03-08 • 13 plant schedules affected

– Interim guidance under consideration to not require 2 plants to accelerate capsule tests

• Requires NRC approval to implement schedule changes – Follow existing protocols for 10CFR50 Appendix H changes – Plant submittals sequenced to distribute review load

• Submission of the request fulfills Needed action – Future change requests will be evaluated by NRC Staff


CRVSP Results • Number of total capsules to be tested by 2025 at or above

the stated fluence

• CRVSP will increase fluence levels of many remaining capsules

Fluence (n/cm2)

Pre-CRVSP

CRVSP

> 3.0x1019 25 29 > 6.0x1019 7 12 > 8.0x1019 1 5 > 9.0x1019 0 2


PWR Supplemental Surveillance Program (PSSP) • A Materials Reliability Program (MRP)-funded research project to

Design, Fabricate & Irradiate 2 supplemental surveillance capsules containing previously-irradiated PWR materials – Reconstitute previously-irradiated specimens (per ASTM

E1253) – Obtain 24 new high-fluence Charpy transition temperature shift

measurements

• Materials selected based on information value to the PWR ETC • Development spread over 3 years, 2012-2014 • Goal: insert capsule(s) in 2014-2015 ; Irradiate ~10 years • Obtain data ~2025 • Flux ~1.2 E+11 n/cm2/s (~0.35 n/cm2/year) • Add ~3.5 E+19 n/cm2 over 10 years • Two irradiation temperatures to more closely match previous

irradiation temperature (~ 539°F, 555°F)


Capsule Design

• Overall geometry fits into standard Westinghouse 3-loop and 4-loop design capsule holder

• Contents – 144 ASTM Type-A Charpy Impact specimens – Dosimetry (Nb, Ni, Fe, Cu, Co) – Melt wire temperature monitors (7 different temps) – SiC as experimental temperature monitor


Materials Selection Method

• Existing surveillance materials were screened for possible selection for PSSP capsules

• Desirability of materials for PSSP is based on – Discrepancy in ETC predictions between current and

potential future ETCs – Ability for these ETCs to predict measured data – The paucity of high-fluence data for the material category

• Materials were broken down into categories based on – Product form (plate, forging, weld) – Cu content – Ni content


Prioritization of Materials for PSSP

Form/Ni Division Cu Group

RG 1.99, Rev. 2 Fit

EONY Fit

WR-C(5), Rev. 1 Fit

Disparity between

ETCs Priority

Plates SA-302BM, SA-533B1

< 0.10 O/U – O Y High 0.10–0.17 –/U – – N Low

> 0.17 – U U N Medium Plates

SA-302B All – – – N Low

Forgings SA-508

< 0.06 –/U –/U O Y High 0.06–0.12 O/U – O Y High

> 0.12 O – – Y High

Welds Ni > 0.4%

< 0.10 O O O Y High 0.10–0.23 – – –/O N Medium

>0.23 O – O Y High

Welds Ni < 0.4%

< 0.10 – – – N Low 0.10–0.23 O/- O O N Medium

>0.23 O - O Y High O = Overpredicts;

U = Underpredicts; – = Reasonable prediction (>1E19 n/cm2) / indicates a difference 3E19 n/cm2


Desired Materials for the Capsule with Lower Irradiation Temperature (Subject to change)

Type Product/Material Group Material ID or

Heat Capsule Number of Specimens

A533B1 Cu < 0.10% B9004-2 X 12 A533B1 0.10% < Cu < 0.17% B6919-1 Z 11 A533B1 Cu > 0.17% B7212-1 V 12 A5082 Cu < 0.06% 125P666 N 12 A5082 0.06% < Cu < 0.12% 125S255 N 12

A5082 0.06% < Cu < 0.12% 123V500 R 12 A5082 Cu > 0.12% 980919/ 281587 Y 12 BOLA Ni > 0.4%, Cu < 0.10% BOLA V 12 L1092 Ni > 0.4%, 0.10% < Cu < 0.23% 1P3571 T 11 L80 Ni > 0.4%, Cu > 0.23% 71249 X 11

L0091 Low Ni, 0.10% < Cu < 0.23% 33A277 Z 11 SMIT

89 Low Ni, Cu > 0.23% 25295 Y 12

A533B1 Low projected USE C5161-1 V 4


Desired Materials for the Capsule with Higher Irradiation Temperature (Subject to change)

Type Product/Material

Group Material ID or

Heat Capsule Number of Specimens

A533B1 Cu < 0.10% R1606-2 W 12 A533B1 Cu < 0.10% B5012-1 W 11 A302B SA-302B PSO102 F 11 A5082 Cu < 0.06% 5P-5933 W 12 A5082 0.06% < Cu < 0.12% 990496/292424 U 11 A5082 Cu > 0.12% 526840 W 11

L80 Ni > 0.4%, Cu < 0.10% 442002 W 11 L124 Ni > 0.4%, Cu < 0.10% 4P4784 Z 12 L124 Ni > 0.4%, Cu < 0.10% PTRO01 V 12 L80 Ni > 0.4%, Cu > 0.23% 61782 N 12 L124 Low Ni, Cu < 0.10% 90209 W 11 L0091 Low Ni, Cu > 0.23% 90136 W83 12 L1092 Low projected USE 20291/12008 W 3

A533B1 Low projected USE C3500-2 Z 3


PSSP Design and Planning – Current Status

• Program makes use of archive irradiated material consistent with NUREG-1801 (NRC’s Generic Aging Lessons Learned report of license renewal)

• Data produced from PSSP will be used by utilities in RV integrity evaluations – In most cases, no anticipated future impact to utility

operation • In process of obtaining permission to use materials for

PSSP capsule from affected utilities – Most materials are not vessel limiting materials and are

unlikely to become limiting • Discussions with prospective host plants have been

initiated


Applicable Guidance of NUREG-1801, Rev 1 (GALL Report) – Retention of Specimens

• “All pulled and tested capsules, unless discarded before August 31, 2000, are placed in storage. (Note: These specimens are saved for future reconstitution use, in case the surveillance program is reestablished.)”


Applicable Guidance of NUREG-1801, Rev 2 (GALL Report) – Retention of Specimens • “[If a plant has ample capsules remaining for future use,….]

Pulled and tested samples, unless discarded before August 31, 2000, and capsules with a neutron fluence greater than 50% of the projected reactor vessel neutron fluence at the end of the period of extended operation are placed in storage (these specimens and capsules are saved for future reconstitution and reinsertion use) unless the applicant has gained NRC approval to discard the pulled and tested samples or capsules.” …and,

• “If an applicant does not have ample capsules remaining for future use, all pulled and tested capsules, unless discarded before August 31, 2000, are placed in storage. (These specimens are saved for future reconstitution use, in case the surveillance program is reestablished.)”


Permissions for Use of Specimens in PSSP

• Specimens are to be retained for possible future reconstitution unless NRC permission to discard the specimens has been received

• The PSSP specimens are not being discarded but are being used for their intended purpose as stated in the GALL Report

• Therefore we conclude that NRC permission to use archive, tested surveillance specimens for the PSSP is not required – However, plant permission will be required – Contributing plants will review relevant plant-specific

commitments


Summary

• Research project to Design/Fabricate/Irradiate (at PWR flux) 2 supplemental surveillance capsules containing ~24 previously-irradiated and reconstituted PWR materials

• Increase fluence up to ~1.2 E+20 n/cm2 to support U.S. PWR operation through 80 years

• Materials selected based on information value to U.S. PWRs – Discrepancy between U.S. ETCs – Paucity of existing and future high-fluence data – Ability to obtain high fluence data through PSSP

• Will ensure high fluence data available for full spectrum of U.S. RPV materials in support of LTO


Environmentally Assisted Fatigue


EAF: EPRI Participating Programs

• EPRI Programs funding EAF Activities – Boiling Water Reactors Vessel and Internals Program

(BWRVIP) – Materials Reliability Program (MRP) – Primary Systems Corrosion Research (PSCR) – Advanced Nuclear Technology (ANT) • Plant Technology – Long-Term Operations (LTO)

• ‘Short-Term’ or Analytical Committee led by Ken Wolfe of BWRVIP

([email protected]) • ‘Long-Term’ or Experimental Committee led by Jean Smith of MRP

([email protected])

mailto:[email protected]�mailto:[email protected]�


EAF Gap Analysis and Roadmap

• Environmentally Assisted Fatigue Gap Analysis and Roadmap for Future Research – Issued December 2012, Product ID: 1026724 – 47 gaps categorized into 7 themes

• 21 gaps identified as high priority


EAF: Priority 1 Roadmap


EAF: Long-Term Experimental Effort

• Objectives – Identify appropriate research that will enhance industry’s

understanding of EAF – Resolve perceived discrepancies between EAF methodology,

existing test data, and industry operating experience – Close the gaps described in the EAF Gap Analysis document

• 10 of the 21 “Priority 1” (high priority) knowledge gaps relate to the

acquisition of data – EPRI collaboration with KHNP to address EAF gaps – 2 “Priority 1” gaps pertain to the development of mechanistic

understanding • EPRI Primary Systems Corrosion Research (PSCR) beginning

efforts in understanding EAF mechanism in 2013/2014


EAF: Long-Term Experimental Effort Tasks Beginning in 2013

• Database to compile published test data on fatigue life • BWR and PWR environments • Carbon & low-alloy steels, austenitic stainless steels, and nickel-based alloys

• Potential comparison of load-controlled and displacement-controlled S-N data • Update MRP-49 Evaluation of Fatigue Data Including Reactor Water

Environmental Effects (2001) with latest data

Task 1: Universal Testing Database (Gap 46)

• Limited data available on the influence of variable temperature and variable strain rate within test cycles and of the influence of out-of-phase variations of temperature and strain rate

• Influence of in-phase and out-of-phase temperature and strain variations to be identified by comparison with isothermal tests

• Austenitic stainless steels in PWR environment are highest priority

Task 2: Non-Isothermal S-N Testing (Gap 15)


EAF: Long-Term Experimental Effort Tasks Beginning in 2013 (cont.)

• No basis for defining the treatment of non-contiguous cycle pairs with regard to both crack initiation and growth

• Pertains to crack initiation in both water environments (PWR & BWR), all materials (CS & LAS, austenitic SS, nickel-based alloys)

• High strain-range cycles separated by long hold times and alternatively separated by long hold times and interjected with a number of low strain range cycles

Task 3: S-N Testing with Complex Transient Loading (Gap 18)

• Basis for the selection of effective stress parameters for biaxial stress conditions is not established

• Axially-loaded pressure tubes subjected to simultaneous axial loading with pressure loading to give non-proportional loading

• Pertains to crack initiation in both water environments and all materials • Comparison to uniaxial data to determine which effective stress parameter gives

the closest correlation for crack nucleation and propagation of microstructurally small cracks

Task 4: Influence of Multiaxial Loading (Gap 36)


EAF: Long-Term Experimental Effort KHNP Research

• Type 316LN and SA 508 Gr. 1a in air, low-DO water, and PWR environment • Fully-reversed, triangular wave form; 3 strain rates; 4 strain ranges

• Alloy 690 in air and PWR environment • Fully-reversed, triangular and mixed wave forms

• Task to be completed 2013

Low-Cycle Fatigue Life of Stainless Steel, Low-Alloy Steel, and Alloy 690

• Work complements Task 3 (S-N Testing with Complex Transient Loading) • Type 316 SS in PWR environment • Varying strain rates and hold times

• 3 different strain rates with 60 second hold time • 3 different strain rates with 300 (or 600) second hold time • Task begins in 2013

Hold-Time Effects Testing of Stainless Steel


EAF: Short-Term Analytical Effort

• Objective: Develop analytical methodology for implementing EAF calculations for license renewal and for Regulatory Guide 1.207

– Utilities are required to evaluate effects of environment on fatigue calculations

– Multiply design “air” fatigue usage by a factor (Fen) that accounts for environment

– Fen can be as large as 20 – Multiplying design fatigue by Fen results in CUF>1 – Design fatigue usage typically includes significant

conservatism – When conservatism is removed CUF is usually


EAF: Short-Term Analytical Effort Task Beginning in 2013

• Primary source of conservatism is the assumed severity of thermal transients (e.g., thermal shock) • Compare EAF analysis with design and realistic transients • Consider high priority locations at several BWRs and PWRs

Task 1: Improved Understanding of Real Plant Transients (Gap 35)

• Builds on method presented to U.S. NRC by BWRVIP/XGEN where a BWR plant example was used to illustrate sufficient margin in design CUF calculations to cover environmental effects for the example component

Demonstrate generically environmental effects are bounded by conservatism inherent in initial design calculations

• List all existing configurations; select most “sensitive” to fatigue • Determine realistic bounding transient (vs. design transient) • Perform one calculation with most sensitive location and bounding transient; demonstrate CUF < 1 • For all plants with similar configurations, EAF is bounded by conservatism in design transients; no

further analysis is required

Demonstrate that for some subset of locations/plants existing fatigue calculations bound environmental effect


EAF: Short-Term Analytical Effort Task 1 Status

• Feasibility study near completion – Components in BWRs and Westinghouse PWRs are

being characterized and binned • Plant specific characteristics being tabulated

– If results show that most components fit into a small number of bins, a likely success path exists • Results to-date are encouraging • Final results June 2013

• Pending acceptable results from feasibility study, detailed calculations will be performed

• Planned completion: 2014


Materials Impact on LTO • Materials Degradation is single biggest threat to nuclear

fleet operations • Materials support for LTO

– Allows owners to develop long-term asset management plans

– Provides tools to support extended operation in a second renewal period

– Provides information needed to make component and station run/replace decisions

– Provides knowledge that supports next generation plants safety and reliability

• Working with LTO to identify gaps and align material efforts to support Long Term Operation


Together…Shaping the Future of Electricity

Materials Research Update for LTOOutlineMaterials Initiative (NEI 03-08) ExpectationsIssue Programs in InitiativeBWR and PWR, Materials Program FocusProgression of Environmental Cracking DamageIntegrated Materials Issues Strategic PlanIndustry Materials Degradation and Issue Management Table ApproachMaterials Degradation Matrix (MDM) Revision 1MDM Revision 1 Strategic Issues2010 MDM RevisionsIdentify Strategic Issues in Materials Degradation MDM Results---- PWR Reactor InternalsMDM Results for PWR: – Neutron Irradiation Effects MDM Results---- BWR Reactor Internals Issue Management Table(s)Issue Management Table(s), cont’dExample IMT – PWRExample Gap Description – PWRPWR IMT UpdateBWR IMT UpdatePWR LTO SummaryBWR LTO SummaryPublication of EPRI MDM and IMTsSlide Number 26BWRVIP Guidelines to Manage DegradationCurrent IMT High Priority GapsCurrent IMT Medium Priority GapsCurrent IMT Low Priority GapsSlide Number 31 MRP ObjectivesKey 2012-13 DeliverablesMRP Guidelines Status2013 Key Research AreasSlide Number 36Irradiated Materials TestingEPRI Irradiated Materials Testing and Degradation Models RoadmapZorita Internals Research ProjectZorita Internals Research Project (ZIRP)Zorita Internals Research Project�Current StatusZorita Internals Research Project�Cutting Plan – Baffle PlatesZorita Internals Research Project�Cutting Plan – Core BarrelMRP Projects Using Additional Zorita MaterialsBWRVIP and PSCR Projects Using Additional Zorita MaterialsHalden Research Project�Halden Research Program�Plant Aging and Degradation�Halden Research Program�2012 to 2014 Test Plan HighlightsHalden Research Program�IASCC: Crack Growth ProgramHalden Research Program�IASCC: Time-to-Failure programIASCC Data Compilation and AnalysisEPRI IASCC Data Compilation & Analysis Objectives and ScopeEPRI IASCC Data Compilation & Analysis�Expert Panel StatusLow-ECP Residuals vs. Average RankEPRI IASCC Data Compilation & Analysis�Next StepsPSCR: Mechanistic understanding on IASCCSlide Number 57Slide Number 58PSCR: Mechanistic Studies on IASCC�IASCC Initiation Mechanism StudyPSCR: Development of Advanced Radiation Resistant Materials (ARRM)PSCR: Development of Advanced Radiation Resistant Materials Program (ARRM)PSCR: Development of ARRMPSCR: Development of ARRMs�Future StepsMRP: Gondole Void Swelling Irradiation and TestingMRP: Gondole Void Swelling Project�Comparison to Void Swelling ModelMRP: Lithium Effects on IASCC InitiationMRP: Effect of Lithium on SCC Initiation �BackgroundMRP: Dynamic Strain Effects on IASCC Initiation RatesMRP: Dynamic Strain Effects on IASCC Initiation Rates �BackgroundMRP: Dynamic Strain Effects on IASCC Initiation Rates Experimental ApproachBWRVIP : Testing of Unirradiated and Irradiated X-750 and XM-19 MaterialsBackgroundX-750 and XM-19 Data NeedsProject StructureSCC CGR Testing - Purpose and Experimental StrategyComparison of GE and INL CGR Tests of �Alloy X-750Preliminary ConclusionsSlide Number 78Background – Alloy 690Alloy 690 PWSCC Knowledge GapsEPRI Alloy 690/52/152 PWSCC Program�Issue StatementAlloy 690/52/152 PWSCC EPRI Program Plan (1/2) Alloy 690/52/152 PWSCC EPRI Program Plan (2/2) CGR TestingSummary Alloy 690 CGR Testing Alloys 52/152 CGR Testing Current StatusPWSCC Initiation Testing for Alloy 690 and Weld MetalsPWSCC Initiation Testing ProjectsSlide Number 89Motivations to Generate LWR Surveillance DataCoordinated Reactor Vessel Surveillance Program (CRVSP)CRVSP ResultsPWR Supplemental Surveillance Program (PSSP)Capsule DesignMaterials Selection MethodPrioritization of Materials for PSSPDesired Materials for the Capsule with Lower Irradiation Temperature (Subject to change)Desired Materials for the Capsule with Higher Irradiation Temperature (Subject to change)PSSP Design and Planning – Current StatusApplicable Guidance of NUREG-1801, Rev 1 (GALL Report) – Retention of SpecimensApplicable Guidance of NUREG-1801, Rev 2 (GALL Report) – Retention of SpecimensPermissions for Use of Specimens in PSSPSummarySlide Number 104EAF: EPRI Participating ProgramsEAF Gap Analysis and RoadmapEAF: Priority 1 RoadmapEAF: Long-Term Experimental EffortEAF: Long-Term Experimental Effort�Tasks Beginning in 2013EAF: Long-Term Experimental Effort�Tasks Beginning in 2013 (cont.)EAF: Long-Term Experimental Effort�KHNP ResearchEAF: Short-Term Analytical EffortEAF: Short-Term Analytical Effort�Task Beginning in 2013EAF: Short-Term Analytical Effort�Task 1 StatusMaterials Impact on LTOTogether…Shaping the Future of Electricity