epri irradiated materials testing programs · 2.1.e. irradiated materials characterization...

Jean Smith Sr. Project Manager

EPRI – NRC Materials Meeting June 5 – 7, 2013

EPRI Irradiated Materials Testing Programs Boiling Water Reactor Vessel Integrity Program

Materials Reliability Program Primary Systems Corrosion Research

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

Contents

• 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


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 PIA 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 (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


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