Ilevbare / Busby / Andersen Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors

1. Auflage 2017
ISBN: 978-3-319-48760-1
Verlag: Springer International Publishing
Format: PDF
Kopierschutz: 1 - PDF Watermark

E-Book, Englisch, 2354 Seiten, eBook

Reihe: The Minerals, Metals & Materials Series

This 15th Edition of the International Conference on Materials Degradation in Light Water Reactors focuses on subject areas critical to the safe and efficient running of nuclear reactor systems through the exchange and discussion of reseach results as well as field operating and management experience.

Zielgruppe

Professional/practitioner

Autoren/Hrsg.

Ilevbare, Gabriell

Busby, Jeremy

Andersen, Peter

Inhaltsverzeichnis

1;Cover;1
2;Title Page;2
3;Copyright Page;3
4;Table of Contents;4
5;Preface;20
6;Conference Organizing Committee/Program Committee;22
7;Session Chairs and Technical Reviewers;23
8;Student Volunteers;25
9;Financial Sponsors;26
10;Opening Session;28
10.1;Overview of NRC Proactive Management of Materials Degradation (PMMD) Program;28
10.1.1;Questions & Answers;37
10.2;Conditions for Long Term Operation of Nuclear Power Plants in Sweden;39
11;Alloy 690 and Its Weld Metals I;51
11.1;Current NRC Perspectives Concerning Primary Water Stress Corrosion Cracking;51
11.1.1;Questions & Answers;65
11.2;Evaluation of the Susceptibility to SCC Initiation of Alloy 690 in Simulated PWR Primary Water;66
11.2.1;Questions & Answers;78
11.3;Role of Cavity Formation on Crack Growth of Cold-Worked Carbon Steel, TT 690 and MA 600 in High Temperature Water;79
11.4;Crack Growth Testing on Cold Worked Alloy 690 in Primary Water Environment;94
11.4.1;Questions & Answers;111
11.5;One-Dimensional Cold Rolling Effects on Stress Corrosion Crack Growth in Alloy 690 Tubing and Plate Materials;112
11.5.1;Questions & Answers;128
12;Alloy 690 and Its Weld Metals II;129
12.1;Cyclic and SCC Behavior of Alloy 690 HAZ in a PWR Environment;129
12.1.1;Questions & Answers;146
12.2;Stress Corrosion Crack Growth Rate Testing of Novel Composite Arrest Specimens;150
12.2.1;Questions & Answers;166
12.3;Environmentally Assisted Crack Growth in Cold Worked Alloy 690TT in Primary Water at Low and High Temperatures;167
12.3.1;Questions & Answers;177
12.4;SCC of Alloy 690 and Its Weld Metals;178
12.4.1;Questions & Answers;194
12.5;SCC Behavior of Alloy 152 Weld in PWR Environment;196
13;Alloy 690 and Its Weld Metals III;214
13.1;Effect of Hot Cracks on EAC Crack Initiation and Growth in Nickel-Base Alloy Weld Metals;214
13.1.1;Questions & Answers;240
13.2;Stress Corrosion Crack Growth of Alloy 52M in Simulated PWR Primary Water;242
13.3;Cyclic and SCC Behavior of Alloy 52M/182 Weld Overlay in a PWR Environment;261
13.4;SCC of High Cr Alloys in BWR Environments;283
13.4.1;Questions & Answers;314
13.5;High-Resolution Characterizations of Grain Boundary Damage and Stress Corrosion Crack Tips in Cold-Rolled Alloy 690;316
14;Alloy 690 and Its Weld Metals IV;330
14.1;Research and Evaluation of Low Temperature Crack Propagation of Ni Base Alloys in Actual Plants;330
14.1.1;Questions & Answers;342
14.2;Penetrative Internal Oxidation from Alloy 690 Surfaces and Stress Corrosion Crack Walls during Exposure to PWR Primary Water;345
14.3;Predicting Chromium Depletion of Nickel-Base Alloys;357
14.3.1;Questions & Answers;369
15;BWR Initiation and Oxide Film Characterization I;371
15.1;In-situ and Ex-situ Oxide Characterization by Synchrotron X-ray (SPring-8) in Non-Sensitized 316 Stainless Steel and High Temperature Water Combination;371
15.1.1;Questions & Answers;383
15.2;High Resolution Electron Microscopy Study on Oxide Films Formed on Nickel-Base Alloys X-750, 182 and 82 in Simulated High Flow Velocity BWR Water Conditions;385
15.2.1;Questions & Answers;398
15.3;Destructive Examinations on Divider Plates from Decommissioned Steam Generators Affected by Superficial Stress Corrosion Cracks;402
15.3.1;Questions & Answers;418
15.4;Non-linear Dynamics of the Morphology at the Oxide / Metal Interface of Austenitic Steels in Simulated Light Water Reactor Environments and Its Implications for SCC Initiation;421
15.4.1;Questions & Answers;432
16;BWR Initiation and Oxide Film Characterization II;436
16.1;The Effect of Cold Work on Microstructure and SCC Susceptibility in Simulated BWR Environment for Non-Sensitized Austenitic Stainless Steels;436
16.2;Behavior of Stress Corrosion Cracking for Type 316L Stainless Steel with Controlled Distribution of Surface Work Hardened Layer in Simulated Boiling Water Reactors Environment;449
16.2.1;Questions & Answers;460
16.3;Influence of Bulk and Surface Cold Work on Crack Initiation and Crack Growth of Austenitic Stainless Steels under Simulated BWR Environment;462
16.3.1;Questions & Answers;478
17;BWR Stainless Steels CGR I;479
17.1;Effect of Nitrogen Addition in 304 L Stainless Steel on the IGSCC Crack Growth Rate in Simulated BWR Environment;479
17.1.1;Questions & Answers;494
17.2;Characterization of Type 304L Stainless Steel: Comparison of ASTM A262 Practice A and Analytical Electron Microscopy Techniques;498
17.2.1;Questions & Answers;508
18;BWR Stainless Steels CGR II;511
18.1;The Effect of Grain Size on IGSCC in SS 316L in Simulated BWR Environment;511
18.1.1;Questions & Answers;524
18.2;An Investigation into Stress Corrosion Cracking of Dissimilar Metal Welds with 304L Stainless Steel and Alloy 82 in High Temperature Pure Water;526
18.2.1;Questions & Answers;535
18.3;Deformation Mode and Microstructure on Stress Corrosion Cracking Path and Kinetics in High Temperature Water Environments;537
18.3.1;Questions & Answers;549
19;Corrosion Fatigue - BWR, PWR;550
19.1;Effect of Static Load Hold Periods on the Corrosion Fatigue Behavior of Austenitic Stainless Steels in Simulated BWR Environments;550
19.2;Effects of Material Composition on Corrosion Fatigue Crack Growth of Austenitic Stainless Steels in High Temperature Water;564
19.2.1;Questions & Answers;583
19.3;Fatigue Limit and Hysteresis Behavior of Type 304l SS in Air and PWR Water, at 150°C and 300°C;584
19.4;NRC Research Activities on Environmentally-Assisted Fatigue;605
19.4.1;Questions & Answers;617
20;Fuel and Fuel Related Materials I;620
20.1;PWR Fuel Deposit Analysis at a B&W Plant with a 24-Month Fuel Cycle;620
20.1.1;Questions & Answers;625
20.2;Effect of DH Concentration on Crud Deposition on Heated Zircaloy-4 in Simulated PWR Primary Water;630
20.2.1;Questions & Answers;642
20.3;Use of AREVA BWR CRUD Model To Study High Zinc Operation at a US Plant;646
20.4;Electrochemical Study of Pre- and Post-Transition Corrosion of Zr Alloys in PWR Coolant;651
20.5;AREVA Fuel Condition Index for a Pressurized Water Reactor;660
21;Fuel and Fuel Related Materials II;665
21.1;Structure and Thermodynamical Properties of Zirconium Hydrides from First-Principle;665
21.1.1;Questions & Answers;674
21.2;Hydride Behavior in Zircaloy-4 during Thermomechanical Cycling;675
22;Fuel and Fuel Related Materials III PWR-BWR;681
22.1;Development of a Method for Studying the Influence of Stress State on the Iodine-Induced Stress Corrosion Cracking of Zirconium Alloys;681
22.2;Wear of Zircaloy-4 Grid Straps Due to Fretting and Periodic Impact with RV Internals Baffle Plates;694
22.2.1;Questions & Answers;707
23;Alloy 718 and X-750;709
23.1;Microstructure and SCC of Alloy X-750;709
23.1.1;Questions & Answers;731
23.2;Stress Corrosion Cracking and Crack Tip Characterization of Alloy X-750 in Boiling Water Reactor Environments;735
23.2.1;Questions & Answers;751
23.3;SCC Properties of Modified Alloy 718 in BWR Plant;753
23.3.1;Questions & Answers;767
23.4;Influence of Chloride Ions as Contaminants on the Corrosion Behavior of Alloy 718 in Pool Water of Nuclear Power Plants;771
24;BWR Low Alloy Steel;781
24.1;Stress Corrosion Cracking Behavior near the Fusion Boundary of Dissimilar Weld Joint with Alloy 182 - A533B Low Alloy Steel;781
24.1.1;Questions & Answers;793
24.2;Effect of Chloride on Environmentally Assisted Cracking of Low Alloy Steels in Oxygenated Hightemperature Water - General Corrosion;795
24.2.1;Questions & Answers;810
25;BWR Stainless Steels CGR III;816
25.1;The Effect of Temperature on the Crack Growth Rate in Simulated BWR Environment;816
25.1.1;Questions & Answers;828
25.2;Effects of Temperature and Corrosion Potential on SCC;837
25.2.1;Questions & Answers;866
25.3;Effect of Thermal Aging on SCC, Material Properties and Fracture Toughness of Stainless Steel Weld Metals;870
25.3.1;Questions & Answers;888
26;Flow Assisted Corrosion;893
26.1;Flow Accelerated Corrosion of Carbon Steel in the Feedwater System of Pwr Plants - Behaviour of Welds and Weld Assemblies;893
26.1.1;Questions & Answers;906
26.2;Modelling Material Effects in Flow-Accelerated Corrosion;910
27;PWR Oxide Films and Characterization;923
27.1;3D Atom-Probe Characterization of Stress and Cold-Work in Stress Corrosion Cracking of 304 Stainless Steel;923
27.1.1;Questions & Answers;931
27.2;Effect of Dissolved Hydrogen, Surface Conditions and Composition on the Electronic Properties of the Oxide Films Formed on Nickel-Base Alloys in PWR Primary Water;936
27.2.1;Questions & Answers;947
27.3;Influence of Primary Water Chemistry on Oxides Formed on Alloy 600 and Alloy690;949
27.4;Characterizing Environmental Degradation in PWRs by 3D FIB Sequential Sectioning;975
27.4.1;Questions & Answers;987
28;PWR Secondary Side/Balance of Plant I;989
28.1;On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes and on Laboratory Specimens;989
28.2;Balance of Plant Corrosion Issues in Aging Nuclear Power Plants;1004
28.2.1;Questions & Answers;1016
28.3;Containment Liner Corrosion;1017
28.3.1;Questions & Answers;1029
29;PWR Secondary Side/Balance of Plant II;1030
29.1;Electrochemical Studies of Steam Generator Tube Degradation in the Presence of Thiosulphate;1030
29.1.1;Questions & Answers;1042
29.2;X-Ray Photoelectron Study of the Oxides Formed on Nickel Metal and Nickel-Chromium 20% Alloy Surfaces under Reducing and Oxidizing Potentials in Basic, Neutral and Acidic Solutions;1044
29.2.1;Questions & Answers;1061
30;SCC of Alloy 82, 182 Welds I;1064
30.1;Interaction of Microstructure, Composition, and Cold Work on the Stress Corrosion Cracking of Alloy 82 Weld Metal;1064
30.1.1;Questions & Answers;1079
30.2;Stress Corrosion Cracking Behavior of Dissimilar Metal Weldments in High Temperature Water Environment;1081
30.2.1;Questions & Answers;1099
30.3;SCC Crack Growth Rate of Alloy 82 in PWR Primary Water Conditions - Effect of a Thermal Treatment;1102
31;SCC of Alloy 82, 182 Welds II;1115
31.1;Initiation of PWSCC of Weld Alloys 182;1115
31.1.1;Questions & Answers;1126
31.2;NRC/EPRI Welding Residual Stress Validation Program (Phase III);1128
31.2.1;Questions & Answers;1139
32;IASCC Stainless Steels CGR I;1141
32.1;Crack Growth Behavior of Irradiated Type 316 SS in Low Dissolved Oxygen Environment;1141
32.1.1;Questions & Answers;1155
32.2;Stress Corrosion Crack Initiation Susceptibility of Irradiated Austenitic Stainless Steels;1159
32.2.1;Questions & Answers;1172
32.3;Stress Corrosion Cracking Behavior of Type 304 Stainless Steel Irradiated under Different Neutron Dose Rates at JMTR;1173
32.3.1;Questions & Answers;1187
32.4;In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR;1188
32.5;Crack Growth Rates of Irradiated Commercial Stainless Steels in BWR and PWR Environments;1198
33;IASCC Stainless Steels CGR II;1210
33.1;The Key Factors Affecting Crack Growth Behavior of Neutron-Irradiated Austenitic Alloys;1210
33.1.1;Questions & Answers;1224
33.2;Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steel WWER Reactor Core Internals;1225
33.2.1;Questions & Answers;1241
33.3;Slow Strain Rate Tensile Tests of Irradiated Austenitic Stainless Steels in Simulated PWR Environment;1244
33.3.1;Questions & Answers;1257
33.4;Irradiation Assisted Stress Corrosion Cracking of Stainless Steels in a PWR Environment (A Combined Approach);1258
33.5;A Preliminary Hybrid Model of Irradiation-Assisted Stress Corrosion Cracking of 300 Series Stainless Steels in PWR Primary Environments;1273
33.5.1;Questions & Answers;1289
34;Irradiation Effects - General I;1292
34.1;Oxidation of a Proton-irradiated 316 Stainless Steel in Simulated BWR NWC Environment;1292
34.2;Irradiation Creep and Irradiation Stress Relaxation of 316 and 304l Stainless Steels in Thermal and Fast Neutron Spectrum Reactors;1302
34.2.1;Questions & Answers;1314
34.3;Recent Insights on the Parametric Dependence of Irradiation Creep of Austenitic Stainless Steels;1318
34.3.1;Questions & Answers;1331
34.4;Cluster Dynamics Prediction of the Microstructure Evolution of 300-series Austenitic Stainless Steel under Irradiation: Influence of Helium;1332
35;Irradiation Effects on Deformation;1344
35.1;Role of Slip Behavior in the Irradiation Assisted Stress Corrosion Cracking in Austenitic Steels;1344
35.1.1;Questions & Answers;1356
35.2;Effect of Environment and Prestrain on IASCC of Austenitic Stainless Steels;1357
35.2.1;Questions & Answers;1373
35.3;Influence of Localized Plasticity on IASCC Sensitivity of Austenitic Stainless Steels under PWR Primary Water;1376
35.4;Deformation Microstructures of 30 dpa AISI 304 Stainless Steel after Monotonic Tensile and Constant Load Autoclave Testing;1387
35.4.1;Questions & Answers;1402
36;PWR Alloy 600 Oxidation and Mechanisms I;1404
36.1;Degradation of Grain Boundary Strength by Oxidation in Alloy 600;1404
36.1.1;Questions & Answers;1418
36.2;Evaluation of the Oxygen Diffusion Coefficient in Nickel-Base Alloys;1419
36.3;Stress Corrosion Cracking of Alloy 600 in PWR Primary Water: Influence of Chromium, Oxygen and Hydrogen Diffusion;1432
36.3.1;Questions & Answers;1444
36.4;Grain Boundary Oxidation and Embrittlement Prior to Crack Initiation in Alloy 600 in PWR Primary Water;1446
37;PWR Alloy 600 Oxidation and Mechanisms II;1457
37.1;Electron Microscopy Characterizations and Atom Probe Tomography of Intergranular Attack in Alloy 600 Exposed to PWR Primary Water;1457
37.1.1;Questions & Answers;1471
37.2;Grain Boundary Oxidation and Stress Corrosion Cracking in Nickel-base Alloys Strained in Supercritical Water;1473
37.2.1;Questions & Answers;1485
37.3;Quantitative Micro-nano (QMN) Approach to SCC Mechanism and Prediction-Starting a Third Meeting;1488
37.3.1;Questions & Answers;1579
38;PWR Alloy 600 SCC I;1582
38.1;Strain Path Effect on IGSCC Initiation and Oxidation of Alloy 182 Exposed to PWR Primary Water;1582
38.1.1;Questions & Answers;1595
38.2;Experimental Study of Short Crack Coalescence in Nickel-Base Alloys In PWR Primary Water;1597
38.2.1;Questions & Answers;1615
38.3;Mechanistic Study on PWSCC of Ni-Based Alloys Using Hump-SSRT Tests under Dry Hydrogen Gas Environment, Simulated Primary Water and Rapid Straining Electrode Test in Simulated Primary Water;1616
38.3.1;Questions & Answers;1632
38.4;Crystallographical Characterization of Initiation of Intergranular Stress Corrosion Cracking of Alloy 600 in PWR Environment;1634
38.4.1;Questions & Answers;1646
38.5;Environmental Effects on PWSCC Initiation and Propagation in Alloy 600;1648
38.5.1;Questions & Answers;1661
39;PWR Alloy 600 SCC II;1662
39.1;Probabilistic Environmentally-Assisted Cracking Modeling for Primary Water Stress Corrosion Cracking of Alloy 600;1662
39.1.1;Questions & Answers;1675
39.2;The Role of Lattice Curvature on the SCC Susceptibility of Alloy 600;1676
39.2.1;Questions & Answers;1689
39.3;PWSCC Susceptibility in Heat-Affected Zones of Alloy 600;1690
39.3.1;Questions & Answers;1700
39.4;Quantitative Residual Strain Analyses on Strain Hardened Nickel Based Alloy;1702
39.5;The Study of Stress Corrosion Cracking on Alloy 600 C-Ring Samples by Polychromatic X-Ray Microdiffraction;1715
39.5.1;Questions & Answers;1725
40;PWR Degradation Management;1728
40.1;Proposed Coordinated U.S. PWR Reactor Vessel Surveillance Program: An Updated Summary Including Program Optimization;1728
40.2;Developing PWR Aging-Management Strategies for Reactor Vessel Internals;1741
40.3;Development of the Extremely Low Probability of Rupture (xLPR) Code;1751
40.3.1;Questions & Answers;1767
40.4;Databases of Operationally Induced Damage;1769
40.4.1;Questions & Answers;1782
41;PWR Water Chemistry and Mitigation;1785
41.1;Introduction to a New Real-Time Water Chemistry Measurement System;1785
41.1.1;Questions & Answers;1792
41.2;Comparison of DBU, NH3, DMA, ETA, and Morpholine Interactions with Ferrous Chloride Solution and Carbon Steel Surfaces;1795
41.2.1;Questions & Answers;1813
41.3;Role of Dissolved Hydrogen in Water in Corrosion of Alloy 600 in High Temperature Water;1815
41.3.1;Questions & Answers;1825
41.4;Quantifying the Benefit of Chemical Mitigation of PWSCC via Zinc Addition or Hydrogen Optimization;1829
41.4.1;Questions & Answers;1843
42;Super Critical Water;1845
42.1;Computational Thermodynamics for Interpreting Oxidation of Structural Materials in Supercritical Water;1845
42.1.1;Questions & Answers;1854
42.2;Stress Corrosion Cracking of Austenitic Alloys in Supercritical Water;1858
42.2.1;Questions & Answers;1872
42.3;Comparison of the Oxidation Behavior of the 14Cr ODS Alloy in Steam and Supercritical Water;1873
42.3.1;Questions & Answers;1882
42.4;Grain Boundary Engineering and Air Oxidation Behavior of Alloy 690;1883
42.4.1;Questions & Answers;1896
43;BWR Water Chemistry and Mitigation I;1897
43.1;Developments in SCC Mitigation by Electrocatalysis;1897
43.1.1;Questions & Answers;1919
43.2;Use of Noble Metal Nanoparticle for SCC Mitigation in BWRs;1922
43.2.1;Questions & Answers;1931
43.3;The Influence of Minor Additions of Platinum Group Metals on Stress Corrosion Cracking in Austenitic Stainless Steels;1932
43.3.1;Questions & Answers;1947
43.4;The Effect of On-Line Noble Metal Addition on the Shut Down Dose Rates of Boiling Water Reactors;1949
43.4.1;Questions & Answers;1963
43.5;Effects of Hydrogen on SCC Growth Rate of Ni Alloys in BWR Water;1964
43.6;Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors;1985
43.6.1;Questions & Answers;2001
44;BWR Water Chemistry and Mitigation II;2002
44.1;Water Chemistry in the Primary Coolant Circuit of a Boiling Water Reactor during Startup Operations;2002
44.2;Simulation of Water Radiolysis by Sonochemistry: Effects on the Electrochemical Behavior of a Stainless Steel;2013
44.2.1;Questions & Answers;2023
44.3;Protective Insulated Coating for SCC Mitigation in BWRs;2024
44.3.1;Questions & Answers;2038
44.4;Influence of Treating Temperature on the Deposition of TiO2 on Type 304 Stainless Steels for Corrosion Mitigation in High Temperature Pure Water;2041
44.4.1;Questions & Answers;2054
45;Irradiation Effects - General II;2056
45.1;Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C;2056
45.1.1;Questions & Answers;2068
46;PWR Field Experience I;2069
46.1;Laboratory Stress Corrosion Cracking Propagation in a Superficial Cold-Worked Layer in SG Divider Plates in Alloy 600;2069
46.2;Destructive Examinations on Divider Plates from Decommissionned Steam Generators Affectined by Superficial Stress Corrosion Cracks;2089
46.2.1;Questions & Answers;2102
46.3;Investigations on Core Basket Bolts from a VVER 440 Power Plant;2105
46.3.1;Questions & Answers;2117
46.4;Laboratory Analysis of a Reactor Coolant Pump Seal;2123
46.5;PWSCC of Thermocoax Pressurizer Heaters in Austenitic Stainless Steel and Remedial Actions to Preventing SCC;2135
47;PWR Field Experience II;2147
47.1;Residual Stress Measurement and the Effect of Heat Treatment in Cladded Control Rod Drive Specimens;2147
47.1.1;Questions & Answers;2160
47.2;Detailed Root Cause Analysis of SG Tube ODSCC Indications within the Tube Sheets of NPP Biblis Unit A;2162
47.2.1;Questions & Answers;2175
47.3;Laboratory Investigation of a Leaking Type 316 Socket Weld in a Boron Injection Tank Sampling Line;2176
47.4;Pressure Tests on SG Pulled Tubes at TSP Level;2188
47.5;Implications of Steam Generator Fouling on the Degradation of Thermal and Material Performance;2200
47.5.1;Questions & Answers;2213
47.6;Key Issues Related to Corrosion Protection of Brackish Water and Sea Water Bearing Components in Cooling Water Systems;2214
48;PWR Initiation and CGR Stainless Steels I;2231
48.1;Understanding the Limits of Lattice Orientation Data Analysis in Environmental Degradation Studies;2231
48.1.1;Questions & Answers;2243
48.2;A Microstructural Investigation on the Effect of Cold Work on Environmentally Assisted Cracking of Austenitic Stainless Steels;2244
48.2.1;Questions & Answers;2256
48.3;Plastic Strain and Residual Stress Distributions in an AISI 304 Stainless Steel BWR Pipe Weld;2260
48.4;Assessment of Lean Grade Duplex Stainless Steels for Nuclear Power Applications;2277
49;PWR Initiation and CGR Stainless Steels II;2289
49.1;Effects of Thermo-Mechanical Treatments on Deformation Behavior and IGSCC Susceptibility of Stainless Steels in PWR Primary Water Chemistry;2289
49.1.1;Questions & Answers;2310
50;Previously Unpublished Manuscripts from the 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems,23-29th August 2009, Virginia Beach, VA, USA;2313
50.1;SCC Initiation Testing of Alloy 600 in High Temperature Water;2313
50.2;The Effects of Deaerated Water on the Toughness of Nickel-based Alloys;2324
50.3;Physical Metallurgy, Weldability, and In-service Performance of Nickel-chromium Filler Metals Used in Nuclear Power Systems;2336
51;Author Index;2347
52;Subject Index;2351

Über Autor(innen)

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