n INFCit,';i.-i,

Proceedings of the

Fifth International Symposium on

ENVIRONMENTAL DEGRADATION OF MATERIALS IN NUCLEAR POWER SYSTEMSWATER REACTORS

August 25-29, 1991 Monterey, California

Sponsored by American Nuclear Society Materials Science and Technology Division The Minerals, Metals, and Materials Society Nuclear Materials Committee National Association of Corrosion Engineers

Published by the American Nuclear Society, Inc. La Grange Park, Illinois 60525 USA

CONTENTS Foreword

XI

Laboratory Verification of On-Line Lithium Analysis Using Ultraviolet Absorption Spectrometry / B. J. Beemster, K. J. Schlager, K. M. Schloegel, S. J. Kahle, T. L. Fredrichs 106

OVERVIEWS

Chairs: D. Cubicciotti (EPRI) and E. P. Simonen (PNL) Materials Degradation Problems in the Advanced Light Water Reactors / T. U. Marston, R. L. Jones Review of Current Research and Understanding of Irradiation-Assisted Stress Corrosion Cracking / J. Lawrence Nelson, Peter L. Andresen

10

The Microstructural and Microchemical Examination of Reactor Pressure Vessel Steels / W. J. Phythian, C. A. 27 English, J. T. Buswell 39

Hydrogen Water Chemistry in BWRs / R. L. Cowan, C. C. Lin, W. J. Marble, C. P. Ruiz

50

59

Discussion

72

118

The Use of Potential Drop Techniques for the Evaluation of Environment Assisted Cracking of Austenitic Alloys / P. Lidar, I. S. Hwang, R. G. Ballinger

126

Discussion

136

Chairs: G. E. Lucas (UCSB) and R. B. Adamson (GE, Pleasanton) 139

Does Silica Play a Role in Zircaloy Corrosion? / V. F. Boston, B. Cheng, C. C. Lin, J. M. Skarpelos 145 An Investigation of Waterside Corrosion of Fuel Cladding in CANDU Reactors Using Infrared Reflection Absorption Spectroscopy / N. Ramasubramanian, M. H. Schankula 150

WATER CHEMISTRY/MONITORING

Chairs: D. D. Macdonald (Penn State Univ) and R. L. Cowan (GE, San Jose)

Influence of Alloying Elements on the Irradiation Hardening and Environmental Sensitivity of Zirconium Alloys / Kjell Pettersson, Lars Hallstadius, Hans Bergqvist, Ake Nylund, Curt Wikstrom 156 77

Corrosion Surveillance Developments at Ontario Hydro / P. E. Doherty, D. C. A. Moore, G. Quirk, A. M. Brennenstuhl 83

Influence of O2 and N2H4 on the ECP in High Temperature Water / B. Stellwag, R. Kilian

In Situ Corrosion Potential Monitoring in Swedish BWRs / Anders Molander, Christer Jansson

Electrochemical Evaluation of Nodular Corrosion Susceptibility of Zircaloy Cladding / Richard A. Perkins..

Zircaloy Performance in Light Water Reactors / R. B. Adamson, B. C. Cheng, R. M. Kruger

Corrosion and Water Chemistry Studies at Halden / T. Karlsen, P. Gunnerud, C. Vitanza

112

ZIRCALOY

Steam Generator Materials Degradation / P. Saint-Paul, G. Slama

An Expert System Assistant to Support the Secondary Side Chemistry Control of a Nuclear Reactor / P. R. Roberge, B. Price, C. M. Daniel, M. J. Dymarski...

SCC Growth Behavior on DCB Specimen of Type 304 Stainless Steel in High Temperature Water / M. Itow, A. Sudo

90 96

Preparation of Zircaloy-4 Specimens with Known Hydrogen Contents / J. H. Zhang, M. Groos, P. Combette, M. Trotabas, M. Labatut, T. Bredel

164

The Characterization of Deuterium Contained in the Protective Oxide of a Zr-2.5 wt% Nb Alloy Using Secondary Ion Mass Spectrometry and Laser Ionization Mass Analysis /A.M. Brennenstuhl, B. D. Warr, N. S. Mclntyre, C. G. Weisener, R. D. Davidson, P. John, G. Smith 169

Dissolution-Passivation Model for Zirconium Alloys in Fluorinated Media / J. Prono, A. Caprani, T. Jaszay, J. P. Frayret 175 A Model for Waterside Oxidation of Zircaloy Fuel Cladding in Pressurized Water Reactors /A. I. A. Amarshad, A. C. Klein

HIGH STRENGTH ALLOYS/ HARD SURFACE/AGING „

Chairs: J. F. Hall (ABB C-E) and J. R. Weeks (BNL) 183 Effect of Heat Treatment on Stress Corrosion of Alloy 718 in Pressurized-Water-Reactor Primary Water / M. T. Miglin, J. V. Monter, C. S. Wade, J. L. Nelson..

Effects of Irradiation on the Microstructural Evolution and Corrosion Resistance of Zirconium. Alloys / J. J. Kai, C. H. Tsai, J. J. Shiao, W. F. Hsieh, C. S. Tu, Y. S. Lee, L. F. Lin, K. Y. Huang

190

Study on a New Zirconium Alloy for Nuclear Reactors / Jianzhang Liu, Peizhi Li

199

Discussion

204

The Effect of Cooling Rate on Precipitate Morphology in Alloy X-750 / M. G. Burke, T. R. Mager, I. L. W. Wilson

287

Modelling of Hydrogen Assisted Cracking of NickelBase Alloy X-750 in Water / T. Oka, R. G. Ballinger, I. S. Hwang 294 Characterization of Long Term Aged Martensitic Stainless Steels / M. Tsubota, K. Hattori, T. Okada

AUSTENITIC DEGRADATION I

279

305

The Evaluation of Iron-Base Hardfacing Alloys on Gate Valves After Cycling Under Simulated PWR Conditions

Chairs: F. P. Ford (GE, Schenectady) and J. C. Danko (Univ of Tennessee)

for One Year / E. V. Murphy, I. Inglis, H. Ocken...

311

APFIM Characterization of the Spinodal Decomposition in Duplex Stainless Steels / J. E. Brown, G. D. W. Smith, P. H. Pumphrey, M. K. Miller 319

Effects of Specific Anionic Impurities on Environmental Cracking of Austenitic Materials in 288°C Water / Peter L. Andresen

209

Effects of Sensitization and Crevices on Critical Cracking Potential for SCC of Alloy 600 / H. Anzai, J. Kuniya, T. Shoji, K. Yoshida

219

Effects of Thermal Aging and Neutron Irradiation on the Mechanical Properties of Stainless Steel Weld Overlay Cladding /EM. Haggag, R. K. Nanstad Discussion

327 333

Crack Propagation in Alloys 600 and 182 in Simulated BWR Environment / Lars G. Ljungberg, Daniel Cubicciotti, Margareta Trolle 226 The Effect of Chromate on IGSCC in Boiling Water Reactors-A SSRT Study / Mats Ullberg

AUSTENITIC DEGRADATION II 235

Stress Corrosion Cracking of 316 SS and Incoloy-800 in High Temperature Aqueous Containing Sulfate and Chloride / Weiguo Zhang, Fangliang Lin, Fengqin Gao, Hongyi Zhou, Xiaoning Cao 240 Stress Corrosion Crack Initiation During Slow Strain Rate Tests on 316 Stainless Steel in Chloride Containing High Temperature Water / W. Yang, J. Congleton, P. Sajdl 245 SCC Growth and Intergranular Corrosion Behavior of Type 316L Stainless Steel in High Temperature Water / Akira Sudo, Mikiro Itow

251

Initiation of Fatigue Crack Growth from Artificial Defects in Forged 316 LNG Stainless Steel / D. Alexander, E. Protopappas, K. D. Bogie

258

Crack Growth of Intergranular Stress Corrosion Cracks — Laboratory and Plant Experience / G. R. Caskey, K. J. Stoner, W. L. Dougherty, R. S. Ondrejcin, R. L. Postles 265 Discussion

272

Chairs: G. M. Gordon (GE, San Jose) and L. G. Ljungberg (ABB-Atom-Sweden) Acoustic Emission from Stress Corrosion Crack Initiation and Growth in Type 304 SS / R. H. Jones, M. A. Friesel

339

An Advanced Coupled Environment Fracture Model for Predicting Crack Growth Rate in LWR Heat Transport Circuits / Digby D. Macdonald, Mima UrquidiMacdonald

345

Prediction of Environmentally Assisted Crack Growh in a Large Diameter Stainless Steel Pipe / S. Ranganath, A. E. Pickett, G. L. Stevens, T. P. Diaz, D. Weinstein, F. P. Ford, R. Pathania 350 Fabrication and Operating History Considerations in Assessing Relative SCC Susceptibility of BWR Components / K. S. Ramp, G. M. Gordon Effect of Surface Preparation on Crack Initiation in Welded Stainless Steel Piping / Joseph C. Danko, Richard E. Smith, David W. Gandy

365

372

Case Study of the Failure of Expansion Bellows in Nuclear Power Plant / /. /. Yang, R.T.Yeh Remediation of Nuclear Plant Piping and Components by Weld Reinforcement Repairs and Weld Buildups / G. J. Licina, A. J. Giannuzzi, J. A. Nevshemal, T. Naughton

STEAM GENERATOR PRIMARY SIDE I 378

Chairs: R. E. Gold (Westinghouse) and G. Slama (Framatome-France) 383

The Role of Carbon and Chromium on the Mechanical and Oxidation Behavior of Nickel-Base Alloys in High Temperature Water / Thomas M. Angeliu, Jin K. Sung, Gary S. Was 475

Weld Repair of Helium Degraded Reactor Vessel Material / W.R. Kanne, Jr., G. J. Bruck, A. Madeyski, D. A. Lohmeier, M. R. Louthan, Jr., D. T. Rankin, R. P. Shogan, G. G. Lessmann, E. A. Franco-Ferreira . . . . 390 Discussion

Grain Boundary Microstructure, Chemistry, and IGSCC in Alloy 600 and Alloy 690 / KJell Norring, Krystyna Stiller, Jan-Olof Nilsson 482

398

The Effect of Grain Boundary Misorientation on Intergranular Cracking of Ni-16Cr-9Fe in 360°C Argon and High Purity Water / Douglas C. Crawford, Gary S. Was.

EROSION CORROSION/SERVICE WATER

488

The Addition of Zinc to Primary Reactor Coolant for Enhanced PWSCC Resistance / J. N. Esposito, G. Economy, W. A. Byers, J. B. Esposito, F. W. Pement, R. J. Jacko, C. A. Bergmann 495

Chairs: A. M. Brennenstuhl (Ontario Hydro-Canada) and C. Jansson (SSPB-Sweden)

Chemistry of Sulfur in High Temperature Water Reduction of Sulfates / B. Sala, P. Combrade, R. Erre, R. Benoit, M. Le Calvar

502

412

Effects of pH and Stress Intensity on Crack Growth Rate in Alloy 600 in Lithiated + Borated Water at High Temperatures / R. B. Rebak, A. R. Mcflree, Z. Szklarska-Smialowska

511

Modelling Hydrodynamic Parameters to Predict Flow Assisted Corrosion / Bryan Poulson, Brian Greenwell, Bindi Chexal, Jeff Horowitz 421

Crack Growth Rate Measurements on Alloy 600 Steam Generator Tubes in Steam and Primary Water / T. B. Cassagne, A. Gelpi

Control of Corrosion Product Transport in PWR Secondary Cycles / S. G. Sawochka, W. L. Pearl, T. O. Passell, C. S. Welty Effects of Absorbed Hydrogen on Flow-Assisted Corrosion in Feedwater Piping: Transient Behavior / Mohammed Abdulsalam, James T. Stanley

405

Primary Water Stress Corrosion Crack Growth Rates in Alloy 600 Steam Generator Tubing / R. G. Lott, R. J. Jacko, R. E. Gold

Erosion-Corrosion and Cavitation-Erosion Measurements on Copper Alloys Utilizing Thin Layer Activation Technique / C. H. Tsai, K. Y. Hsu, J. J. Kai, T. N. Yang

430

Clad Piping—A Novel Approach for Solving Nuclear Plant Service Water and Erosion-Corrosion Problems / Bhaven Chakravarti

436

442

456

Chemical Decontamination Solutions: Effects on PWR Equipment / C. M. Pezze, E. R. Colvin, R. G. Aspden

462

Discussion

470

539

LOW ALLOY STEELS/EAC

Materials Problems and Enhancements for LWR Heat Exchangers in Taiwan / J. I. Lee, H. C. Lai, J. P. Yang, W. Y. Mao, F. Chu, G. J. Tseng, B. T. Fan, H. N. Horng 450 Fouling and Corrosion of Freshwater Heat Exchangers / A. M. Brennenstuhl, T. S. Gendron, P. E. Doherty..

525

The Effects of Chemical Factors on Stress Corrosion of Alloy 600 Exposed to the Cooling Medium in Pressurized Water Reactors / Ph. Berge, F. de Keroulas, O. Menet, A. Rocher, P. Pitner, A. Gelpi, G. Pinard-Legry.. 533 Discussion

Material Selection and Life Extension Techniques to Mitigate the Degradation of Service Water System Components at Susquehanna Nuclear Power Plant / Raymond S. Tombaugh, Jack R. Maurer

518

Chairs: P. Scott (Framatome-France) and T. Shoji (Tohoku Univ-Japan) Determination of the Threshold Values for Corrosion Fatigue Crack Growth Rate of Pressure Vessel Steels in PWR Primary Water / H. E. Hannxnen, E. Arilahti, U. Ehrnste'n 545 Crack Tip Conditions Related to Environmentally Assisted Cracking in Pressure Vessel Steels: Effect of Temperature / P. Combrade, M. Foucault vu

554

Stress Corrosion Cracking of Low-Alloy Steels in High Temperature Water / F. P. Ford, P. L. Andresen, D. Weinstein, S. Ranganath, R. Pathania

PRESSURE VESSEL RADIATION 561

Chairs: G. R. Caskey (Westinghouse SRC) and L. M. Davies (Nuclear Electric-UK)

Studies of Stress Corrosion Cracking in Steels Used for Reactor Pressure Vessels / W. A. Van Der Sluys, R. Pathania 571 Stress Corrosion of Low Alloy Steels Used in External Bolting on Pressurised Water Reactors / P. Skeldon, N. R. Smart, P. Hurst

Investigations of Irradiation-Anneal-Reirradiation (IAR) Property Trends of RPV Welds / J. R. Hawthorne, A. L. Hiser 671

579

SANS Investigation of Low Alloy Steels in Neutron Irradiated, Annealed, and Reirradiated Conditions / R. Kampmann, F. Frisius, H. Hackbarth, P. A. Beaven, R. Wagner, J. R. Hawthorne 679

Investigation of Service-Induced Degradation of Steam Generator Shell Materials / W.H. Bamford, G. V. Rao, J. L. Houtman 588

Characterization of Copper Precipitation in a 17/4 pH Steel: A Combined APFIM/TEM Study / M. K. Miller, M. G. Burke 689

Life Prediction for Nuclear LP Rotor SCC Cracks / J. Y. Liu, E. E. Lai, C. C. Su, H. C. Lai, David H. R. Lin 596 Discussion

Effects of Composition and Temperature on Irradiation Hardening of Pressure Vessel Steels / G. E. Lucas, G. R. Odette, E. Mader, F. Haggag, R. Nanstad

604

Use of Miniature and Standard Specimens to Evaluate Effects of Irradiation Temperature on Pressure Vessel Steels /EM. Haggag, R. K. Nanstad, S. T. Byrne ..

STEAM GENERATOR PRIMARY SIDE II

The Influence of Dissolved Hydrogen on Primary Water Stress Corrosion Cracking of Alloy 600 at PWR Steam Generator Operating Temperatures / R. J. Jacko, G. Economy, F. W. Pement

Measurement of Irradiation Damage in Nuclear Pressure Vessel Steels by Magnetic Properties Change / J.F. Stubbins, J. G. Williams, A. M. Ougouag, J. U. Patel, W.-J. Shong 719

609

Discussion

725

613 STEAM GENERATOR SECONDARY SIDE I

The Resistance to PWSCC of Explosively Expanded Alloy 600 Tube-to-Tubesheet Joints / R. E. Gold, F. W. Pement, S. A. Tarabek, G. Economy 621 Examination of a Steam Generator Tube Removed from Maine Yankee / T. P. Magee, P. J. Plante

704

Multivariable Modeling of Reactor Pressure Vessel Weld Fracture Toughness / E. D. Eason, J. E. Wright, E. E. Nelson 711

Chairs: J. P. N. Paine (EPRI) and J. Stubbe (Laborelec-Belgium) Correlation of Temperature with Steam Generator Tube Corrosion Experience / J. A. Gorman, R. A. Ogren, J. P. N. Paine

696

Chairs: G. P. Airey (CEGB-UK) and A. Mcllree (EPRI)

628 Selection of Statistical Distributions for Prediction of Steam Generator Tube Degradation / K. D. Stavropoulos, J. A. Gorman, R. W. Staehle, C. S. Welty, Jr. ..

Leak Rate and Burst Test Data for McGuire Unit 1 Steam Generator Tubes / P. A. Sherburne, C. R. Frye, D. B. Mayes 636

Predicting Steam Generator Crevice Chemistry / J. P. N. Paine, S. A. Hobart, S. G. Sawochka

Recent Observations in Steam Generators at the Bruce Nuclear Power Development / Peter J. King, Francisco Gonzalez 644

731 739

A Detailed Model of Localized Concentration Processes in Porous Deposits of SG's / Peter J. Millett, James M. Fenton 745

Cracking of Alloy 600 Heater Sleeves and Nozzles in PWR Pressurizers / J. F. Hall, D. B. Scott, D. A. Wright, R. S. Pathania

652

Stress Corrosion Cracking of Pressurizer Instrumentation Nozzles in the French 1300 MWe Units / D. Alter, Y. Robin, M. Pichon, A. Teissier, B. Thomeret

Evaluation of SG Crevice Environment by Directly Sampled Method Using an On-Site Autoclave Facility / H. Takamatsu, K. Matsueda, E. Kadokami, K. Arioka, T. Tsuruta, S. Okamoto, T. Ueno 752

661

Accelerated IGA/SCC Testing of Alloy 600 in Contaminated PWR Environments / B. P. Miglin, J. M. Sarver, D. W. Koch, K. Aoki, H. Takamatsu

Discussion

667 vni

757

Lead-Induced Stress Corrosion Cracking of Alloy 600 and 690 in High Temperature Water / T. Sakai, T. Senjuh, K. Aoki, T. Shigemitsu, Y. Kishi 764 Straining Electrode Behavior and Corrosion Resistance of Nickel Base Alloys in High Temperature Acidic Solution / Kazuo Yamanaka

773

IGA/IGSCC of Alloy 600 in Acidic Sulfate and Chloride Solutions / W.H. Cullen, M. J. Partridge, J. A. Gorman, J. P. N. Paine

780

Discussion

789

IRRADIATION ASSISTED STRESS CORROSION CRACKING I

Chairs: P. L. Andresen (GE, Schenectady) and F. Garzarolli (Siemens/KWU-FRG) Irradiation-Induced Sensitization and Stress Corrosion Cracking of Type 304 Stainless Steel Core-Internal Components / H. M. Chung, W. E. Ruther, J. E. Sanecki, T. F. Kassner Stress Corrosion Cracking of High Energy ProtonIrradiated Stainless Steels / J. M. Cookson, R. D. Carter, D. L. Damcott, M. Atzmon, G. S. Was, P. L. Andresen An IASCC Study Using High Energy Ion Irradiation / K. Fukuya, K. Nakata, A. Horie Irradiation-Induced Chromium Depletion and Its Influence on Intergranular Stress Corrosion Cracking of Stainless Steels / S. M. Bruemmer, L. A. Chariot, E. P. Simonen

IGA Resistance of TT Alloy 690 and Concentration Behavior of Broached Egg Crate Tube Support Configuration / S. Suzuki, T. Kusakabe, H. Yamambto, K. Arioka, T. Ochi

861

Assessment of Autogenous Type 41 OS Stainless Steel Welds in Replacement Steam Generator Tube Support Structures / J. M. Sarver, R. R. Seeley, M. D. Lees..

869

Corrosion Fatigue of Alloy N06600 Steam Generator Tubing / Gabriel I. Ogundele, Peter J. King

877

Electrochemical Behavior of Alloy 800 Under Steam Generators Temperature and Pressure Conditions: Comparison Based on (I,E) Polarization Curves, with 316L and A533B Steels / T. Jaszay, J. P. Frayret, A. Caprani

885

Is OD Peening of Alloy 800 SG Tubes Desirable? / J. Stubbe, F. Wolters, P. Somville

893

Effectiveness of 700°C Thermal Treatment on Primary Water Stress Corrosion Sensitivity of Alloy 600 Steam Generator Tubes: Laboratory Tests and In Field Experience / F. Cattant, D. Garriga-Majo, F. de Keroulas, P. Todeschini, J. C. van Duysen 901

795

Discussion

.'

912

806

IRRADIATION ASSISTED STRESS CORROSION CRACKING II

814

Chairs: G. S. Was (Univ of Michigan) and S. M. Bruemmer (PNL) 821

High-Temperature Solution Annealing as an IASCC Mitigation Technique / A. J. Jacobs, C. M. Shepherd, G. E. C. Bell, G. P. Wozadlo

917

827

Calculation of Corrosion Potentials in Boiling Water Reactors / Digby D. Macdonald

935

832

Investigation of the Protection Potential Against IASCC / M. E. Indig, J. Lawrence Nelson, G. P. Wozadlo

941

Changes in Grain Boundary Composition Induced by Neutron Irradiation on Austenitic Stainless Steels / K. Asano, K. Fukuya, K. Nakata, M. Kodama

838

Effects of Fluence and Dissolved Oxygen on IASCC in Austenitic Stainless Steels / M. Kodama, S. Nishimura, J. Morisawa, S. Suzuki, S. Shima, M. Yamamoto ...

948

Discussion

844

Stress Corrosion Crack Growth Rate of Sensitized Type 304 Stainless Steel During High Flux Gamma-Ray Irradiation in 288°C Water / K. Nakata, S. Shimanuki, H. Anzai, K. Mabuchi, M. Fuse, N. Shigenaka

955

End-of-Life Irradiation Performance of Core Structural Components in the Shippingport Light Water Breeder Reactor / J. C. Clayton, B. C. Smith

963

Discussion

970

Attendees

973

Author Index

979

Subject Index

983

Grain Boundary Segregation and Intergranular Stress Corrosion Cracking Susceptibility of Austenitic Stainless Steels in High Temperature Water / T. Shoji, K. Yamaki, R. G. Ballinger, I. S. Hwang Irradiation Assisted Degradation of Grain Boundaries in an Fe-Cr-Ni Alloy / S. Yamamoto, Y. Ishida, N. Sekimura

STEAM GENERATOR SECONDARY SIDE II

Chairs: B. Miglin (B&W) and H. Takamatsu (Kansai Electric-Japan) The Temperature Dependence of the Tensile Properties of Thermally Treated Alloy 690 Tubing / D. L. Harrod, R. E. Gold, B. Larsson, G. Bjoerkman Thermal Treatment, Grain Boundary Composition and Intergranular Attack Resistance of Alloy 690 / A. J. Smith, R. P. Stratton

849

855

IX