Astm stp 870 1985

J., "In-Reactor Creep of Selected Ferritic Alloys," Effects of Radiation on Materials: Twelfth International Symposium, ASTM STP 870, F.. For irradiation temperatures less than 520°C, t

ASTM SPECIAL TECHNICAL PUBLICATION 870

F A Garner, Westinghouse Hanford Co and

J S Perrin, Office of Nuclear Waste Isolation, editors

ASTM Publication Code Number (PCN) 04-870000-35

1916 Race Street, Philadelphia, PA 19103

Library of Congress Cataloging-in-Pub!ication Data

Effects of radiation on materials

(ASTM STP; 870)

Papers presented at the Twelfth International

Symposium on the Effects of Radiation on Materials

"ASTM publication code number (PCN) 04-870000-35."

Includes bibliographies and index

1 Materials—Effect of radiation on—Congresses

I Garner, F A II Perrin, J S III ASTM

Committee E-10 on Nuclear Technology and Applications

IV International Symposium on Effects of Radiation on

Materials (12th; 1984: Williamsburg, Va.) V Series:

ASTM special technical publication; 870

TA418.6.E333 1985 620.1'1228 85-11257

ISBN 0-8031-0450-2

Copyright © by AMERICAN SOCIETY FOR T E S T I N G A N D M A T E R I A L S 1985

Library of Congress Catalog Card Number; 85-11257

N O T E The Society is not responsible, as a body, for the statements and opinions advanced in this publication

Printed in Baltimore, MD November 1985

Foreword

The symposium on Effects of Radiation on Materials: Twelfth International

Symposium contains papers presented at the Twelfth International

Sympo-sium on the Effects of Radiation on Materials The sympoSympo-sium was sponsored

by ASTM Committee E-10 on Nuclear Technology and Applications J S Perrin, Office of Nuclear Waste Isolation, presided as chairman with F A Garner, Westinghouse Hanford Company, and J J Koziol, Combustion Engineering, Inc., as cochairmen J S Perrin and F A Garner are editors of this publication

Related ASTM Publications

Effects of Radiation on Materials (11th Conference), STP 782 (1982), 04-782000-35

Effects of Radiation on Materials (10th Conference), STP 725 (1981), 04-725000-35

Effects of Radiation on Structural Materials (9th Conference), STP 683 (1979), 04-683000-35

Effects of Radiation on Structure and Mechanical Properties of Metal, STP

529 (1973), 04-529000-35

A Note of Appreciation

to Reviewers

The quality of the papers that appear in this publication reflects not only the obvious efforts of the authors but also the unheralded, though essential, work of the reviewers On behalf of ASTM we acknowledge with appreciation their dedication to high professional standards and their sacrifice of time and effort

ASTM Committee on Publications

ASTM Editorial Staff

Helen M Hoersch Janet R Schroeder Kathleen A Greene Bill Benzing

Contents

Overview 1

IRRADIATION CREEP OF STRUCTURAL METALS

In-Reactor Creep of Selected Ferritic Alloys—RAYMOND J PUIGH 7

Evaluation of Ferritic Alloy Fe-lViCr-lMo After Neutron Irradiation:

Irradiation Creep and Swelling—DAVID S GELLES AND

RAYMOND J PUIGH 19

Influence of a Temperature Change on In-Reactor Creep—

BRYAN A CHIN AND E ROBERT GILBERT 38

Non-Isothermal In-Reactor Creep of Nickel Alloys Inconel 706 and

P E - 1 6 — E ROBERT GILBERT AND BRYAN A CHIN 5 2

In-Pile Creep Strain and Failure of Cold-Worked Type 316 Titanium

Pressurized Tubes—JEAN-LOUIS BOUTARD,

ARLETTE MAILLARD, YVETTE CARTERET, VIVIANE LEVY, AND

JEAN-MARIE BOYER 6 1

Discussion 74

Critical Assessment of Low-Fluence Irradiation Creep Mechanisms—

CHARLES H HENAGER, JR., AND EDWARD P SIMONEN 7 5

Irradiation-Creep-Induced Anisotropy in a/1 (110) Dislocation

Populations—DAVID S GELLES 98

MiCROSTRUCTURAL DEVELOPMENT

Effect of Irradiation Temperature on the Precipitation in Cold-Worked

Titanium-Stabilized Type 316 Stainless Steel—DIDIER GILBON,

LUCIEN LE NAOUR, CHRISTIAN RIVERA, AND HENRI LORANT 115

Microstructure of Irradiated Inconel 706 Fuel Pin Cladding—

WALTER J S YANG AND BRUCE J MAKENAS 127

Microsegregation Observed in Fe-35.5Ni-7.5Cr Irradiated in

EBR-II—HOWARD R BRAGER AND FRANK A GARNER 139

Transmission Electron Microscope Studies and Microhardness Testing

of Irradiated Ferritic Steels—DEBORAH K H U L E T T A N D

WILLIAM A JESSER 151

Phase Stability in Irradiated Alloys by Constrained Equilibrium

Thermodynamics—JAMES PAUL HOLLOWAY AND

J A M E S F STUBBINS 1 6 7

N E U T R O N - I N D U C E D S W E L L I N G

Swelling of Austenitic Iron-Nickel-Chromium Ternary Alloys During

Fast Neutron Irradiation—FRANK A GARNER A N D

HOWARD R BRAGER 187

Swelling of 20% Cold-Worked Type 316 Stainless Steel Fuel Pin

Cladding and D u c t s — B R U C E J M A K E N A S 202

Swelling of AISI Type 304L Stainless Steel in Response to

Simultaneous Variations in Stress and Displacement Rate—

DOUGLAS L PORTER AND FRANK A, GARNER 212

Some Observations on the Effect of Stress on Irradiation-Induced

Swelling in AISI Type 316 Stainless Steel—THOR LAURITZEN,

WALTER L BELL, JERRY M ROSA, AND SAM VAIDYANATHAN 221

Swelling of Microstructure of Neutron-Irradiated Titanium-Modified

Type 316 Stainless S t e e l — J E A N LOUIS SERAN,

LUCIEN LE NAOUR, PIERRE GROSJEAN, MARIE PIERRE HUGON,

YVETTE CARTERET, AND ARLETTE MAILLARD 233

Role of Dislocations, Dislocation Walls, and Grain Boundaries in Void

Formation During Early Stages of Fast Neutron Irradiation—

ANDY HORSEWELL AND BACHU N SINGH 248

Discussion 260

C H A R G E D P A R T I C L E I R R A D I A T I O N

Influence of Applied Stress on Swelling Behavior in Type 304 Stainless

Steel —NAOHIRO IGATA, YUTAKA KOHNO, HIDEO TSUNAKAWA,

AND TATSUHIKO FUJIHIRA 265

Stainless Steel and Fe-20Ni-15Cr Alloy—AKIRA KOHYAMA,

BEN A LOOMIS, GUY AYRAULT, AND NAOHIRO IGATA 277

Discussion 296

Experimental Investigation of the Effect of Injected Interstitials on

Void Formation—BUCKY B A D G E R , J R , D O N A L D L PLUMTON,

STEVEN J ZINKLE, ROBERT L SINDELAR,

GERALD L KULCINSKI, RICHARD A DODD, AND

WILHELM G WOLFER 297

Influence of Helium on Swelling of Steels—VIVIANE LEVY,

DIDIER GILBON, AND CHRISTIAN RIVERA 317

Comparison of Depth-Dependent Microstructures of Ion-Irradiated

Type 316 Stainless Steels—ROBERT L SINDELAR,

R ARTHUR DODD, AND GERALD L KULCINSKI 330

Experimental Determination of the Critical Cavity Radius in Fe-lOCr

for Ion Irradiation—LINDA L H O R T O N A N D L O U I S K M A N S U R 344

Discussion 357

Comparison of Thermal and Irradiated Behavior of High-Strength,

High-Conductivity Copper Alloys—STEVEN J ZINKLE,

R ARTHUR DODD, AND GERALD L KULCINSKI 363

Ion Bombardment Damage in a Modified Fe-9Cr-lMo Steel—

KENNETH FARRELL AND EAL H LEE 383

Helium and Displacement Damage Produced by 600 MeV Proton

Beams in High Purity Aluminum—DIDIER GAVILLET,

ROLF G O T T H A R D T , JEAN-LUC MARTIN, SHERRON L GREEN,

WALTER V GREEN, AND MAXIMO VICTORIA 394

Direct Observation of Cascade Defect Formation at Low Temperatures

in Ion-Irradiated M e t a l s — T A K E O M U R O G A , K O I C H I HIROOKA,

AND SHIORI ISHINO 407

Solute Segregation and Void Formation on Grain Boundaries in

Electron-Irradiated Type 316 Stainless Steel—SOMEi OHNUKi,

HEISHICHIRO TAKAHASHI, AND TARO TAKEYAMA 419

Nonequilibrium Segregation and Phase Instability in Alloy Films

During Elevated-Temperature Irradiation in a High-Voltage

Electron Microscope—NGHI Q LAM AND PAUL R OKAMOTO 430

THEORY OF SWELLING

Modeling of Void Swelling in Irradiated Steels—BRUCE B GLASGOW

AND WILHELM G WOLFBR 453

Influence of Composition on Steady-State Void Nucleation in

Irradiated Alloys—BAHRAM ESMAILZADEH AND

ARVIND S KUMAR 468

Effect of Microstructure on the Minimum Critical Radius and Critical

Number of Gas Atoms for Swelling—WILLIAM A COGHLAN

AND LOUIS K MANSUR 481

Discussion 492

Dual-Ion Irradiation: Impact of the Conflicting Roles of Helium on

Void Nucleation—ARVIND S KUMAR AND FRANK A GARNER 493

Discussion 505

Nucleation of Voids—The Impurity Effect—i-WEl CHEN AND

ADEMOLA TAIWO 507

A Mechanism of Void Lattice Formation Based on Two-Dimensional

Self-Interstitial Diffusion—JOHN H EVANS 525

Index (see Volume II) 1245

is not included since this topic receives adequate treatment in other forums

In the first section a significant amount of recent data on Irradiation Creep

of Structural Metals is compiled, concentrating on both ferritic and austenitic

alloys Continuing a trend established earlier in this symposium series, there are several papers detailing the creep response of alloys to nonisothermal reactor histories Two papers in this section concentrate on the microstruc-tural origins of irradiation creep, with one demonstrating conclusively that creep deformation generates and is in turn sustained by the development of

an anisotropic distribution of dislocation Burgers vectors

The second section on Microstructural Development explores the changes

in microstructure, microcomposition, and phase stability that accompany and cause macroscopic changes in physical properties or dimensions A sig-nificant new observation is that irradiation can induce spinodal decom-position in some alloys which are not known to decompose by this mechanism

in nonradiation environments

In Neutron-Induced Swelling a large amount of data is shown that

demon-strates that austenitic alloys tend to swell at a rate of ~ l % / d p a following a transient regime While this posttransient regime is very insensitive to compo-sition, temperature, and other variables, the duration of the transient regime

is quite sensitive to major element composition, particularly at relatively high irradiation temperatures Minor elements such as titanium also exhibit

a pronounced influence There is a minimum transient period, however, of

—10 dpa that cannot be shortened by variations in composition or mental variables One new and surprising conclusion is that the application

environ-of compressive stress does not lengthen the transient duration as has been

1

2 EFFECTS OF RADIATION ON MATERIALS

routinely assumed Both compressive and tensile stresses were shown to equally influence void nucleation so as only to shorten the transient regime

In the section on Charged Particle Irradiation the application of tensile

stresses was confirmed to operate on void nucleation and the transient tion, but a correspondence between void and Frank loop development was also found A number of papers explored the influence of helium on radiation-induced microstructural development, although some differences of opinion are expressed as to whether the influence of helium observed in the simulation will be representative of that experienced in the neutron environment Several papers in this and the next section explore the possibility that the injected interstitial represented by the bombarding ion not only distorts the swelling response but also the effect of helium One significant finding presented in this section is that highly focused electron beams cause segregation of alloy components and thereby create phase instabilities in a manner quite atypical

dura-of the neutron environment Charged particle irradiation was also used to cast that high-strength copper alloys may undergo a significant degradation

fore-in mechanical properties as a result of radiation-affected dislocation recovery and grain recrystallization

In Theory of Swelling most papers focus on the nucleation stage of void

formation, exploring the role of impurities, composition, and helium One paper proposes a model for the formation of void lattices based on two-dimensional diffusion of self-interstitials and the shadowing effect of voids

on the diffusion of interstitials in their vicinity

The Mechanical Properties section contains a wide variety of papers One

group of papers explores the radiation-induced changes in fracture toughness

of iron-based austenitic and ferritic alloys as well as that of various conium alloys Another group of papers addresses fatigue behavior in ther-mal and fast reactors, while yet another group considers the microstructural origin of radiation-induced mechanical property changes The data pre-sented in the symposium both confirmed and extended the prevailing percep-tions of the effect of radiation on mechanical properties; no significantly new

zir-or different phenomena were disclosed, however

The Pressure Vessel Steels section of this conference has been growing at

each meeting A review was made of irradiation testing performed over the last 12 years on pressure vessel steels and their weldments The studies of the last few years have confirmed the significance of chemistry control in govern-ing irradiation resistance of ferritic steels Copper has consistently shown up

as the principal element over which control must be maintained However, other elements such as nickel, when combined with copper, enhance copper's effect, and phosphorus and manganese were shown by some investigators to have a measurable influence The discernment of the actual mechanisms in-volved for each of these elements is not made easy due to the inability to date

to see clear evidence of radiation damage using electron microscopy Field ion emission microscopy may offer promise over conventional transmission

electron microscopy The thought was also offered that boron is of cance due to its transmutation where thermal to fast neutron ratios are high Although the chemistry issue is not totally resolved, current damage trend curves do show a definitive relationship to the major contributory element, copper, especially at high fluences The current status appears to be that the irradiation effects data have been exhaustively analyzed and correlation with postulated damage models is reasonably good Further work is necessary to provide microstructural evidence of damage and establish relatively narrow bounds on alloy composition The concentration of experiments on com-mercial pressure vessel alloys has not provided the range and variety of ele-ments necessary to confidently establish these bounds, however The concen-tration of effort on Charpy tests was also questioned and examination of other properties related to mechanical behavior, including microhardness changes, has been undertaken The relation of the damage to fracture tough-ness has been more closely studied, but the significance of the upper shelf energy values provided by Charpy curves still escapes full understanding The annealing-out of radiation damage was reviewed in a number of papers The likelihood of successfully "wet" annealing of pressure vessel damage for extended times at 650°F was shown to be relatively low A higher temperature "dry" anneal is considered feasible but its cost-effectiveness and the treatment of the vessel nozzles and their attached piping remain as concerns

signifi-In the Irradiation Facilities section only two of the papers presented were

provided for these proceedings These papers involve the Los Alamos Meson Physics Facility (LAMPF) The absent papers addressed the proposed Fu-sion Materials Irradiation Test (FMIT) Facility and the Materials Open Test Assembly (MOTA), an experimental test facility currently operating in the FFTF fast reactor in Richland, Washington

The final section on Other Radiation Studies covers three papers that do

not easily fit in the other categories These are a modified method of helium introduction into alloys via the tritium trick, radiation damage aspects of a novel method for providing safe storage of Krypton-85 from fuel re-processing, and radiation effects on resins and zeolites forming part of the waste stream from the clean-up effort at the Three Mile Island Nuclear Plant

F A Gamer

Hanford Engineering Development tory, Richland, WA 99352; symposium co- chairman and coeditor

Irradiation Creep of Structural Metals

Raymond J Puigh^

In-Reactor Creep of Selected

Ferritic Alloys

REFERENCE: Puigh, R J., "In-Reactor Creep of Selected Ferritic Alloys," Effects of

Radiation on Materials: Twelfth International Symposium, ASTM STP 870, F A

Garner and J S, Perrin, Eds., American Society for Testing and Materials, phia, 1985, pp 7-18

Philadel-ABSTRACT: In-reactor creep data to a peak fluence of 5.7 X lO" neutrons (n)/cm^

( £ > 0.1 MeV) for irradiation temperatures between 380 and 570°C have been tained for the ferritic alloys HT9, 9Cr-lMo, and 2y4Cr-lMo Also, in-reactor creep

ob-data for HT9 has been obtained at 6$0°C at a fluence of 2.9 X lO" n / c m ' (£ > 0.1

MeV) For irradiation temperatures less than 520°C, the creep strains for all ferritic loys are small and less than predicted for 20% cold-worked Type 316 stainless steel At 570°C, thermal creep mechanisms become important and 2y4Cr-lMo loses its creep resistance HT9 loses its creep resistance at 650°C and exhibits large diametral strains (>24%)

al-KEY WORDS: in-reactor creep, ferritic alloys, irradiation-induced creep, radiation

The long-term structural integrity of any large fission breeder or fusion energy device is required for the commercial viability of such a reactor The ferritic alloy class has been shown to exhibit several desirable properties when considering them for application in such devices For example, a number of commercial ferritic alloys have shown a high resistance to irradia-tion-induced swelling to fluences of 1.8 X lO" neutrons (n)/cm^ ( £ > 0.1

MeV) (90 displacements per atom (dpa)) [1] However, microstructural

ex-aminations of these alloys [2] have shown that secondary phases are formed during neutron irradiation in several ferritic alloys that may impact their mechanical properties

Recent in-reactor creep results on selected ferritic alloys [5] to a peak ence of 56 dpaF (French dpa or 37 dpa) for the temperature range 400 to 500°C at hoop stresses of 205 and 290 MPa show small in-reactor creep

flu-strains Low dose creep data from the Mol 5B experiment [4] suggests a

lin-ear fluence dependence for the in-reactor creep behavior of the Martensitic Steel No 1.4914 after an initial rapid transient Data on HT9 to a peak flu-ence of 1.0 X lO" n/cm^ (£• > 0.1 MeV) also indicate a linear fluence de-

' Senior scientist, Westinghouse Hanford Company, Richland, WA 99352

7

8 EFFECTS OF RADIATION ON MATERIALS

pendence and a stress dependence with a stress exponent of approximately 1.5 [5]

The present experiment was undertaken to quantify the in-reactor creep behavior of selected ferritic alloys as part of the U.S National Clad/Duct Materials Development Program The creep data for HT9 and 2/4Cr-lMo

from the first discharge of this experiment were reported in Ref 6 This

re-port covers the results from the second discharge of this experiment for diation temperatures between 380 and 570°C to a peak fluence of 5.7 X 10 n/cm^ (£• > 0.1 MeV) (28.5 dpa) This report also includes creep data for the ferritic alloy 9Cr-lMo and creep data at 650°C on the ferritic alloy HT9 at a

irra-fluence of 2.9 X lO'' n/cm^ {E > 0.1 MeV) (14 dpa)

Experimental Procedure

Pressurized tube creep specimens were fabricated from the ferritic alloys HT9, 9Cr-lMo, and 2y4Cr-lMo, using techniques described in Ref 7 The chemical composition and thermomechanical treatment for these alloys are given in Table 1 These alloys were received in bar stock form and drilled to produce tubing 4.57 mm outside diameter by 4.17 mm inside diameter End-caps fabricated from HT9 were electron beam welded to tubing segments 19.81 mm in length This geometry was chosen to ensure an adequate wall thickness of 0.2 mm and yet optimize the use of the limited irradiation vol-ume One endcap had a capillary hole for pressurization of the specimen Each specimen was filled with helium to the desired pressure, and the closure weld for gas containment was made with a laser beam that passed through the glass port of the pressure vessel and sealed the capillary fill hole in the endcap All specimens were helium leak checked prior to irradiation Also, definitions for the parameters describing the mechanical state of the pressur-

ized tube creep specimens are given elsewhere [8] The specimen diameters

were measured using a noncontacting laser system [9] that has an accuracy

of ±2.5 X 10'" mm Measurements were performed at five equidistant tions (0.267-cm apart) symmetric about the midplane of the specimen The middle three measurements were averaged to yield an average diameter for the specimen The maximum diameter measurement of the five measure-ments was used to calculate the maximum diametral strain The hoop strain for a given specimen was determined from measurement of its diameter be-fore and after irradiation The repeatability in the hoop strain measurement

posi-is 0.05%

The irradiation of these specimens occurred in the Experimental Breeder Reactor (EBR-II) at Idaho Falls The irradiation vehicle used in the first ir-

radiation of these specimens at 400, 450, and 550°C is described in Ref 6

These specimens were reconstituted into three different irradiation vehicles that were cylindrical tubes 1.5 m in length and 2.0 cm in diameter The

"weeper" irradiation vehicle had holes in the cylinder's wall that directly

10 EFFECTS OF RADIATION ON MATERIALS

posed the specimens to the EBR-II sodium coolant The other two tion vehicles were designed for irradiation temperatures of 450 and 550°C The specimens in these two irradiation vehicles were loaded into isothermal subcapsules filled with sodium-potassium (NaK) that positioned the speci-mens axially within the core region The dimension of the insulating gas gap between the subcapsule and the outer cylinder was designed to control the heat transferred from the gamma-heated subcapsule to the reactor coolant and thus to yield the desired irradiation temperatures The data at 650°C were obtained from pressurized tube creep specimens irradiated in a "heat pipe" capsule that provides compensatory temperature control by transport-ing heat from the higher to the lower temperature portions of the capsule by means of vaporization and condensation of the liquid metal vapor [9] The irradiation temperatures for the creep specimens were determined

irradia-from thermal expansion difference devices (TED) [10] located with the

spec-imens Two TED's were located with the specimens in each irradiation

vehi-cle The TED analysis for the first irradiation experiment is given in Ref 6

The peak temperatures for the second irradiation of these specimens are given in Table 2 Also given in Table 2 are the analyses of the TED's in the heat pipe irradiation vehicle These temperatures in Table 2 have been cor-rected for measured volume changes in the TED cladding material The av-erage temperatures in Table 2 for the second irradiation were averaged with the TED measurements from the previous irradiation to obtain the average irradiation temperatures for the creep data reported in this paper These av-erage irradiation temperatures were 419, 490, 572, and 653°C, respectively The irradiation of the pressurized tube creep specimens at 419, 490, and 572°C occurred in a Row 4 position of EBR-II The fluence values for the creep specimens were estimated from the known axial flux profile normal-ized to the estimated peak fluence The estimated peak fluence was deter-mined by multiplying the peak flux for a Row 4 position, 1.75 X 10'^ n/cmVs (£ > 0.1 MeV), by the time during which the irradiation vehicle was

in the reactor (4636 h) The uncertainty in the fluence assignments is

esti-TABLE 2—TED temperature assignments

IR17 IR13 IR12 IR20 IR18 IR9

PI P3

Calculated TED Temperature,

395 ± 12

475 + 7

570 ± 7

653 ± 1

mated to be ±10% based upon analyses performed on dosimetry

experi-ments performed in EBR-II A similar procedure was used to assign fluences

to the creep specimens irradiated in the "heat pipe" irradiation vehicle that was irradiated in a Row 2 position in EBR-II The peak flux for such a posi-tion is 1.85 X 10" n/cmVs (£• > 0.1 MeV) The "heat pipe" irradiation ve-hicle was in the reactor for 4477 h before the specimens were examined

Experimental Results

The diametral change data from this examination of the in-reactor creep experiment's specimens are given in Tables 3 to 5 Table 4 also contains data from the first examination that have not previously been reported These ta-bles list the specimen identification, total fluence, the time at temperature, the average irradiation temperature, hoop stress, diametral strain, and effec-tive creep strain [7] for each specimen Creep data from the previous dis-

charge of this experiment may be found in Ref 6 The 2V4Cr-lMo alloy

exhib-its the largest diametral change (0.09%) in the zero-stressed specimens at 419°C In general, the diametral changes in the zero-stress specimens for all the alloys are qualitatively largest for the average radiation temperature of 419°C At 419 and 490°C, the creep behaviors of all the alloys are similar

TABLE 3—HT9 (91353) in-reactor creep data

5,2 5.2 5.2 5.2 5.7 5.7 5.7 5.7 5.2 5.2 5.2 5.2 2.9 2.9 2.9

e,

%

0.0 0.13 0.21 0.37 0.0 0.15 0.32 0.39 0.0 0.26 0.43 0.66 0.0 0.50 1.61

12 EFFECTS OF RADIATION ON MATERIALS

TABLE 4—9Cr-7Mo (91887) in-reactor creep data

2.7 2.7 2.7 2.7 5.6 5.6 5.6 5.6 2.7 2.7 2.7 2.7 5.6 5.6 5.6 5.6 1.8 1.8 1.8 1.8 4.7 4.7 4.7 4.7

e

%

0.0 0.04 0.05 0,08 0.0 0.06 0.11 0.16 0.0 0.04 0.06 0.10 0.0 0.10 0.11 0.18 0.0 0.13 0.29 0.40 0.0 0.18 0.34 0.54

TABLE 5—2'hCr-lMo (38649) in-reactor creep data

5.0 5.0 5.0 5.0 5.7 5.7 5.7 5.7 5.4 5.4 5.4 5.4

€ ,

%

0.0 0.14 0.16 0.23 0.0 0.05 0.14 0.18 0.0 0.95 2.41 3.01

"Specimen ruptured

with HT9 to exhibiting shghtly higher creep strains when compared to IMo and 2V4Cr-lMo At 570°C, the 2y4Cr-lMo alloy exhibits large in-reac-tor creep strains when compared to the other ferritic alloys in the experi-ment At 570°C, the measured diametral strain in the 100 MPa stress specimen for 2/4Cr-lMo (Specimen PJ65) shows no change from the mea-sured diametral strain for this examination and its previously measured strain, which suggests that the specimen failed sometime during the first ir-radiation period At 653°C, the HT9 alloy exhibits large creep strains with an average diametral strain of 8.7% in the HT9 Specimen NR14 (45 MPa) However, the maximum strain measured in NR14 was 24.5% This mea-surement suggests that the 60 MPa hoop stress specimen (NR15) for HT9 has failed for the 653°C irradiation (The maximum diametral strain for Speci-men NR15 was 17.3%.)

9Cr-The in-reactor creep data for these ferritic alloys at 419 and 490°C indicate that the stress exponent describing the in-reactor creep stress dependence is between 1 and 2 At 570°C, the HT9 and 9Cr-lMo alloys continue to show a similar stress dependence, and the 2V4Cr-lMo alloy data indicate a stress ex-ponent of the order 2 to 6 At 653°C, the HT9 in-reactor creep data suggest a stress exponent of the order 3 to 7 The creep data for these ferritic alloys are consistent with a linear fluence dependence for irradiation temperatures less than 510°C

Discussion

The effective creep strain data for these ferritic alloys are shown as a tion of hoop stress for the different irradiation temperatures in Figs 1 to 4 Also shown in these figures is the calculated in-reactor creep behavior of a

func-1 func-1 func-1 ^ / ' > ^ ' IRRADIATION TEMPERATURE; ^S'C / jT

HOOP STRESS MPa

FIG 1—Comparison of ferritic creep data at 419^ C v/ith predictions for the in-reactor creep of

20% cold-worked Type 316 stainless steel given in Ref 11

14 EFFECTS OF RADIATION ON MATERIALS

" V I 1 / IRRADIATION TEMPERATURE: 490°C

O HT9

O 9Cr-1Mo

n 2 1/4Cr-1Mo 316SS CORRELATION +1o

DATA UNCERTAINTY

HOOP STRESS, MPa

FIG 2—Comparison offerritic creep data at 49ff'C with predictions for the in-reactor creep of

20% cold-worked Type 316 stainless steel given in Ref 11

specific heat of cold-worked AISI Type 316 stainless steel [7i] The solid curve is the nominal prediction of the correlation, and the dashed curves are estimated standard deviation limits on the correlation At 419 and 490°C, all four ferritic alloys exhibit less creep strain when compared to the Type 316 stainless steel correlation The difference between the Type 316 stainless steel correlation predictions and the ferritic data is larger for this examination of

the creep specimens than at the lower fluence examination [6\ Several

Q 2 1/4Cr-1Mo ' y

/ /

/ /n

/ / / /

^ / ^ / / ^^''

—

1

HOOP STRESS, MPa

FIG 3—Comparison offerritic creep data at 57(fC with predictions for the in-reactor creep of

20% cold-worked Type 316 stainless steel given in Ref 11

IRRADIATION TEMPERATURE: 653°C

O HT9

316SS CORRELATION

±1o

HOOP STRESS, MPa

FIG 4—Comparison offerritic creep data at 653° C with predictions for the in-reactor creep of

20% cold-worked Type 316 stainless steel given in Ref 1 \

tenitic stainless steels [72] exhibit a nonlinear fluence dependence that are consistent with climb-induced glide mechanisms The ferritic steels investi-gated in this work, however, exhibit a linear fluence dependence Therefore, the percentage difference in the creep strains between austenitic steels and ferritic steels could increase with increasing neutron fluence At 570°C, the Type 316 stainless steel correlation predicts significantly more creep strain than the strains measured for the ferritic alloys HT9 and 9Cr-IMo The mea-sured creep strains in 2V4Cr-lMo are comparable to the predicted creep strains for Type 316 stainless steel At 653°C, the measured creep strains for HT9 are significantly larger than the predicted creep strains for Type 316 stainless steel Therefore, for temperatures greater than approximately 640°C, cold-worked Type 316 stainless steel offers superior creep resistance when compared to the ferritic steels investigated in this experiment

To investigate the temperature dependence of the ferritic creep behavior, the following in-reactor creep model has been assumed

where e is the effective creep strain (%), 4)t is the fast fluence (10^^ n/cm^),

a is the effective stress (MPa), and n is the stress exponent The average creep

coefficient, B, contains the temperature dependence for the in-reactor creep

behavior The in-reactor creep data from this experiment and the creep data

from Ref 6 are consistent with a stress exponent of 1 < « < 2 Therefore, in our analyses, we have chosen n = 1.5 The data from this experiment (Tables

3 to 5 and Ref 6) were used to calculate for each alloy an average B for a

16 EFFECTS OF RADIATION ON MATERIALS

given fluence and irradiation temperature A plot of 5 as a function of diation temperature is shown in Fig 5 The standard deviation in the averages are indicated by the error bars whenever they are larger than the symbol size

irra-The open symbols represent values for B calculated from creep data from the

first discharge of the experiment where the fast fluence was less than

2.8 X 10^^ n/cm^ (£ > 0.1 MeV) The solid symbols represent values for B

calculated from creep data from the second discharge of the experiment For

irradiation temperatures less than 520°C, the values for B for a given ferritic

alloy remain constant One interpretation of these results is that the average creep coefficient for ferritic alloys is relatively insensitive to temperatures in the range 380 to 520°C and that the in-reactor creep data has a linear fluenc£

dependence for fluences to 5.7 X 10^' n/cm' (E > 0.1 MeV) At 570°C, the B

values for all ferritic alloys investigated are larger Based upon limited

ther-mal creep data discussed in Ref 6, these larger values for B are attributed to

" 30

in

-r T" T"

Ot < 3.0 X 1022 0 t > 4.7x1022 n/cm2 n/cm2

FIG 5—The temperature dependence for the ferritic alloys' average creep coefficients, B, are

shown for this experiment and the data from Ref 6

the onset of thermal creep mechanisms Finally, at 653°C, the Bvalues for

HT9 become more than an order of magnitude larger than their B values at

570°C At 650°C, the thermal creep mechanisms are assumed to dominate the in-reactor creep behavior of this ferritic alloy

When comparing the relative ranking of the alloys' creep resistance (Fig 5), HT9 exhibits an average creep coefficient that is approximately a factor

of two larger than the other ferritic alloys investigated for irradiation temperatures less than 505°C At 570°C, 2y4Cr-lMo loses its in-reactor creep

resistance Also^at 570°C, the values for B at the lower fluence appear to be larger than the B values calculated from the second discharge data If the 5 values were normalized to time at temperatures instead effluence, then the B

values from the two discharges would be similar, which is consistent with the interpretation that thermal creep mechanism are^important at this tempera-

ture Another interpretation of the difference in B values at 570° C for

differ-ent fluence could be that at this temperature the ferritic alloys have a cant primary creep component

signifi-The B value for 2y4Cr-lMo is qualitatively higher at the higher fluence and

lower temperature The volumetric change from the zero-stressed specimen

was 0.3% assuming isotopic volume change Therefore, the higher B values

for 2y4Cr-lMo at 419°C could be attributed to a swelling-enhanced creep mechanism seen in several austenitic steels Higher fluence data, however, would be needed to corroborate such a hypothesis

Finally, the in-reactor creep data at 653°C for HT9 suggest that this alloy

no longer displays the in-reactor creep resistance it exhibited at irradiation temperatures less than 570°C The measured average creep strains for HT9 are much larger than predicted for 20% cold-worked Type 316 stainless steel [77] More importantly, HT9 exhibits relatively large local ductilities before failure The average and maximum diametral strains measured at 653°C are given in Table 3 for HT9 These data suggest that localized diametral strains

in excess of 24% are possible for this alloy

Conclusions

All the ferritic alloys investigated in this experiment continue to exhibit superior in-reactor creep resistance when compared to Type 316 stainless steel for irradiation temperatures less than 520°C, and their average creep coefficients are independent of temperature and fluence At 570°C, the aver-age creep coefficients for all ferritic alloys are larger, indicative of the onset

of thermal creep mechanisms At 570°C, the 2 /4Cr-lMo ferritic alloy appears

to have lost its creep resistance for the stress levels investigated At 570°C, the HT9 and 9Cr-lMo alloys continue to exhibit creep strains significantly smaller than those predicted for 20% cold-worked Type 316 stainless steel [77] Finally, at 653°C, the HT9 ferritic alloy has lost much of its creep strength Evidence for significant localized diametral strain (>24%) was ob-served) in the 45 MPa hoop stress specimen (NR14) condition at 653°C

18 EFFECTS OF RADIATION ON MATERIALS

References

[/] Powell, R W,, Peterson, D X, Zimmerschied, M K., and Bates, J F., Journal of Nuclear

Materials, Vols 103 and 104, 1981, pp 969-974

[2] Gelles, D S., Journal of Nuclear Materials, Vols 103 and 104, 1981, pp 975-980

[3] DeBremaecken, A and Huet, J J in Proceedings, Conference on Dimensional Stability

and Mechanical Behaviour of Irradiated Metals and Alloys, Vol 1, Brighton, UK, 11-13 April 1983, pp 117-120

[4] Herschback, K and Doser, W in Proceedings, Conference on Dimensional Stability and

Mechanical Behavior of Irradiated Metals and Alloys, Vol 1, Brighton, UK, 11-13 April

1983, pp 121-124

[J] Straalsund, J L and Gelles, D S., "Assessment of the Performance Potential of the tensitic Alloy HT9 for Liquid Metal Fast Breeder Reactor Applications;" and Chin, B A.,

Mar-"An Analysis of Creep Properties of a 12Cr-lMo-W-V Steel," Proceedings, Topical

Con-ference on Ferritic Steels for Use in Nuclear Energy Technologies, Snowbird, UT, 19-23, June 1983, in press

[5] Puigh, R J and Wire, G., "In-Reactor Creep of Selected Ferritic Alloys," Proceedings,

Topical Conference on Ferritic Steels for Use in Nuclear Energy Technologies, Snowbird,

UT, 19-23, June 1983, in press

[7] Gilbert, E R and Chin, B A., Nuclear Technology, Vol 52, Feb 1981, pp 273-283 [81 Gilbert, E R and Blackburn, L D., Journal of Engineering Materials and Technology, Vol

99, April 1977, pp 168-180

[P] Gilbert, E R and Chin, B A in Effects of Radiation on Materials: Tenth Conference,

ASTM STP 725, D Kramer, H R Brager, and J S Perrin, Eds., American Society for

Testing and Materials, Philadelphia, 1980, pp 665-679

[/O] Franklin, D C and Reuther, D S., Transactions, American Nuclear Society, Vol 14, Nov

1971, p 632

[ i / ] Puigh, R J., Gilbert, E R., and Chin, B A in Effects of Radiation on Materials: 11th

Inter-national Symposium, ASTM STP 782, H R Brager and J S Perrin, Eds., American

So-ciety for Testing and Materials, Philadelphia, 1982, pp 108-121

[12] Schneider, W., Herschback, K., and Ehrlich, K in Effects of Radiation on Materials—11th International Symposium, ASTM STP 872, H R Brager and J S Perrin, Eds., American

Society for Testing and Materials, Philadelphia, 1982, pp 30-43

David S Gelles^ and Raymond J Puigh^

Evaluation of Ferritic Alloy

Fe-2V4Cr-1Mo After Neutron

Irradiation: Irradiation Creep

and Swelling

REFERENCE: Gelles, D S and Puigh, R J., "Evaluation of Ferritic Alloy

Fe-2V4Cr-IMo After Neutron Irradiation: Irradiation Creep and Swelling," Effects of Radiation on

Materials: Twelfth International Symposium, ASTM STP 870, F A Garner and J S

Perrin, Eds., American Society for Testing and Materials, Philadelphia, 1985,

pp 19-37

ABSTRACT: Irradiation creep and swelling measurements are reported for

Fe-2y4Cr-IMo after irradiation by fast neutrons over the temperature range 390 to 560''C ameter change measurements on thin-walled pressurized tubes in a bainitic condition and density change measurements on rods in a nonstandard condition were made fol- lowing irradiation in the Experimental Breeder Reactor II The irradiation creep spec-

Di-imens were irradiated toafluence of 5.7 X 10" neutrons (n)/cm^ (E > 0.1 MeV)or 30

displacements per atom (dpa) and the swelling specimens were irradiated to a peak fluence of 2.4 X 10" n/cm or 115 dpa These results have been used as a basis to es- tablish in-reactor creep and swelling correlations for 2/iCr-lMo in a bainitic condi- tion The correlations predict moderate swelling and moderate irradiation enhanced creep at 390°C The in-reactor creep at 570°C indicates that thermal creep mechanisms become important at this irradiation temperature

KEY WORDS: irradiation creep, thermal creep, swelling, swelling-enhanced creep,

design correlations, swelling equation, in-reactor creep equation, radiation, Fe-2V4Cr-lMo

The first wall and blanket structure are key components in a fusion tor, since the structural life can impact the reactor's performance that ulti-mately translates into the cost of electricity To date, a number of metals have been proposed for possible use as the structural material in the first wall and blanket, and many have been included in conceptual power reactor de-signs These conceptual designs have helped to show that while each material has attractive properties or features for use in design (low thermal stresses, high temperature strength, or low, long-term radioactivity), they also have disadvantages and not one of these materials including Type 316 stainless

reac-' Principal engineer and senior scientist, respectively, Westinghouse Hanford pany, Richland, WA 99352

Com-19

2 0 EFFECTS OF RADIATION ON MATERIALS

Steel has emerged as the overwhelming favorite for use in commercial tors The primary uncertainty regarding the use of these materials is a lack of radiation damage information, particularly with respect to the synergism be-tween neutron damage and transmuted helium In an effort to gain a better understanding of these effects and to ultimately develop a radiation-resistant material, the Office of Fusion Energy (OFE) of the Department of Energy (DOE) created, in 1976, a program known as Alloy Development for Irradi-ation Performance (ADIP)

reac-At the time the ADIP program was created, there were more than ten ferent alloy systems proposed for use in the first wall and blanket structure Rather than study all of these materials, it was decided to group them into families or classes and emphasize only the most promising material within each group These groups are refeii d to as paths and there are ci rrently five paths in the ADIP program These are Path A—Austenitic Stainless Steel [Type 316 and modified 316 called prime candidate alloy (PCA)]; Path B— iron-nickel-chromium precipitation strengthened alloys (625, X750, PE-16); Path C—Reactive and Refractory Metals (niobium, vanadium, and titanium alloys); Path D—Innovative Materials and Concepts (long ranged ordered alloys, ceramics, composites); and Path E—Ferritic Steels (HT-9, 9Cr-lMo, and 2V4Cr-lMo)

dif-The primary objective of the ADIP program is to develop materials ble of operating in a fusion reactor up to a time integrated neutron exposure

capa-of 40 MW year/m^ or 500 displacements per atom (dpa) A secondary tive is to provide both materials and design data for use in lower perfor-mance, intermediate fusion systems such as ETR and DEMO In the parallel path approach, each of the candidates are independently brought through a series of sequential steps consisting of scoping studies, base research studies, and alloy optimization This type of approach is necessary since premature selection or rejection of an alloy could severely limit the design options available in the future Currently, the only system to be in the alloy optimiza-tion stage is Type 316 stainless steel

objec-Since the inception of the ADIP program, the primary emphasis has been

on irradiation experiments with the net result that roughly 7000 specimens have been or are currently being irradiated The specimens consist of tensile, creep-rupture, fatigue, fracture-toughness, and flaw-growth specimens as well as pressurized tubes and transmission electron microscopy (TEM) disks The specimens cover the whole range of alloys including 20% cold-worked (CW) Type 316 stainless steel, ferritic steels, nickel alloys, titanium alloys, vanadium alloys, and long-range ordered alloys Because of the large number of specimens involved and a limited fusion budget, only a small frac-tion of these specimens have been tested The remaining specimens have been archived for future testing when more funds become available or testing priorities change Currently, the bulk of the ADIP effort is being directed towards modified Type 316 stainless steel (PCA) and ferritic steels (HT-9 and

9Cr-lMo) By initiating a program to test other candidate materials such as Fe-2/4Cr-lMo steel, the data base for irradiated material can be substan-tially improved, which will enable improved material development planning, life prediction, and material selection

Ferritic steels, particularly those containing 9 to 13% chromium are of terest to the Liquid Metal Fast Breeder Reactor (LMFBR) program for use

in-in claddin-ing and ducts because of the alloy's elevated temperature strength, creep resistance, compatibility with a liquid metal coolant, and availability from industry For the same reasons that these steels are of interest to the breeder program, they are also of interest to the fusion reactor program for use in the first wall and blanket structure However, for fusion reactor appli-cations, ferritic steels offer specific advantages over Type 316 stainless steel, namely, low thermal stresses because of their better thermal conductivity and lower thermal expansion While ferritic steels will have roughly twice the thermal stress resistance of Type 316 stainless steels, this resistance will be lower than for the other candidate materials such as the vanadium alloys The biggest advantage with the ferritic alloys rests in its resistance to radia-tion damage Experiments conducted on ferritic steels as part of the breeder reactor National Cladding/Duct Materials Development program revealed that, as a class, these materials are more resistant to void swelling and irradi-ation creep than Type 316 stainless steel

While these initial scoping irradiations indicate that the ferritic steels have the potential for increased component lifetimes in comparison to Type 316 stainless steel, more information is needed about their resistance to neutron irradiation Of particular concern is the shift in the ductile-to-brittle transi-tion temperature and the swelling resistance at higher fluences To increase the understanding of the behavior of ferritic steels to neutron irradiation, the breeder program in 1975 and the fusion program in 1979 initiated a number

of irradiation experiments designed to provide information on swelling, creep, tensile, and fracture toughness behavior of a number of ferritic steels The primary emphasis of these experiments was to examine ferritic steels capable of operating at temperatures <550°C such as 12Cr-lMo (HT-9) and 9Cr-lMo; however, the lower strength steel 2V4Cr-lMo was also included be-cause of its widespread unirradiated application in the chemical process and nuclear industries

Currently, the 2 /4Cr-lMo steel has not been seriously considered for use in either fusion or fission components because of its limited creep strength at temperatures above about 500°C As a result, the irradiated 2V4Cr-lMo spec-imens have not been tested and DOE has no current plans to include them in their post-irradiation experiments However, recent reactor studies such as STARFIRE and MARS indicate that acceptable electrical power genera-tion can be achieved with first walls operating at temperatures <500°C For temperatures <500°C, the 2V4Cr-lMo steel in the normalized and tempered condition has properties equivalent to the 12Cr-lMo steel and, as a result

22 EFFECTS OF RADIATION ON MATERIALS

becomes a viable alternative An additional advantage is in its better ability The high chromium-molybdenum steels such as 12Cr-lMo are more sensitive to cold cracking in weld heat-affected zones than the lower chro-mium steels such as 2y4Cr-lMo For the large complicated structures used in fusion, which will likely require a number of welds, improved weldability can

weld-be a distinct advantage Even though the 2y4Cr-lMo steel appears to offer an advantage over the other ferritic steels based on fabricability and at lower temperatures has equivalent properties, little is known about its radiation resistance The present effort is designed to provide this information so that its potential can be further explored

Objectives and Technical Approach

The objectives of this ferritic steel study are to develop an understanding

of the response of the 2V4Cr-lMo steel to neutron irradiation and to present the results of the mechanical property evaluations and swelling studies in a format consistent with its inclusion in the Materials Handbook for Fusion Energy Systems (MHFES) The results of these evaluations will be inter-preted in terms of their impact on future studies of this alloy and its suitabil-ity for use in fusion reactor components

Experimental Procedures

Materials for irradiation creep and swelling experiments were obtained from different sources The creep specimens were fabricated from a Mannes-mann Heat 38649 provided by Climax Molybdenum Company and swelling specimens were fabricated from a Lukens Steel Company sample, Heat C4337-14S (also identified as Alloy A-387-D) The compositions, as supplied

by the vendors, are provided in Table 1 The Mannesmann heat was received

TABLE 1—Chemical analysis of 2'/4Cr-lMo heats

as supplied by the vendors (in percent by weight)

Swelling Rods*

0.12 0.42

0.21 0.16 2.17 0.93

balance

"Mannesmann Company Heat 38649

'Lukens Steel Company Heat C4337-14S

2 1/4Cr-1Mo 0.75"

T ROLL(1150°C)

1 0.507"

T HT-1: 760°C/lh, RAC

I

COLD ROLL (12%)

0.446"

T

STRAIGHTEN (950°C)

I MACHINE 0.25" RODS

I MACHINE TUBES

FIG I—Rolling schedule for Alloy I'UCr-lMo used for AAXIV

in the form of a 12.7-cm-long section of 7-cm wall pipe The section was cut radially in 2-cm-thick slices, which were subsequently rolled and machined into tubes according to the rolling schedule diagrammed in Fig 1 Tube di-mensions were 0.457 cm outside diameter by 0.417 cm inside diameter The specimens were heat treated according to the schedule given in Table 2 End-caps of HT-9, a martensitic stainless steel, were electron beam welded to tub-ing segments 1.981 cm in length This geometry was chosen to ensure an adequate wall thickness of 0.02 cm and yet optimize the use of the limited ir-radiation volume One endcap had a capillary hole for pressurization Each specimen was filled with helium to the desired pressure and the closure weld

TABLE 2—Heat treatments given creep tubes

and swelling specimens prior to irradiation

Specimen Type Heat Number Heat Treatment"

Creep tubes 38649 900°C/30 min/AC +700°C/1 h/AC

Swelling specimens C4337-14S 1010°C/1 h/WQ + 843''C/2 h/WQ

"Temperature/time at temperature/cooling procedure where AC = air cooled and

WQ = water quenched

24 EFFECTS OF RADIATION ON MATERIALS

for gas containment was made with a laser beam that passed through the glass port of the pressure vessel and sealed the capillary fill hole in the end-cap Specimen diameters were measured both before and after irradiation using a non-contacting laser system that has an accuracy of ±2.5 X 10^^ cm and has repeatability in the hoop strain measurement of 0.05% The Lukens Steel Company heat was sectioned into random cross sections approximately

1 cm in diameter and heat treated according to the schedule given in Table 2

The somewhat unusual heat treatment was based on information from Ref 1

Specimens 0.3-cm diameter by 1.3-cm long were then machined from the stock Swelling was determined from density measurements based on the Archimedian principle Multiple measurements were made on each specimen with a typical measurement uncertainty of ±0.05%

Specimens were irradiated in the Experimental Breeder Reactor (EBR-11) located in Idaho Falls The irradiation vehicle for the creep specimens, iden-tified as Capsule B329 and part of the AAXIV experiment, was a cylindrical tube 1.5 m in length and 2.0 cm in diameter Inside were three subcapsules that were connected to the sodium coolant flow (called a "weeper" design)

by a capillary tube with an inlet at the bottom of the capsule and an outlet at the top of the capsule The sodium flow was necessary to achieve the desired lowest temperature in the capsule and permit gas release from the subcapsule should a creep specimen rupture The dimensions of the insulating gas gap between the subcapsule and the outer capsule was designed to control the heat transferred from the gamma heated subcapsule to the reactor coolant that was flowing past the outer capsule wall Calculations were performed to optimize the sodium flow rate through the capsules so as to minimize the thermal gradient within a given subcapsule The nominal design tempera-tures for each subcapsule were 400, 450, and 550°C Capsule B329 was loaded into Subassembly X359 that was irradiated in Position 4C2 in EBR-II for Cycles 109 through 111 and 113 The specimens were in the reactor for a period of 10680 megawatt days (MWD), which corresponds to 4477 h at temperature and a peak fluence exposure 2.8 X 10^^ neutrons (n)/cm^ (£• > 0.1 MeV) or 14dpa The reconstitution of the AAXIV experiment that contained the ferritic creep specimens consisted of three separate B7 cap-sules Capsules B331, B333, and B334 contained the ferritic pressurized tube specimens reconstituted from the AAXIV three-temperature capsule (B329) and were designed for the irradiation temperatures of 550, 450, and 400°C, respectively Capsule B334 was a weeper design and, therefore, the speci-mens were directly exposed to the sodium coolant These B7 capsules were part of Subassembly X359a that was irradiated in Position 4C2 in EBR-II for Cycles 116 through 119 The specimens were in the reactor for a period cor-responding to 10979 MWD that corresponds to 4603 h at temperature and a peak fluence exposure of 2.9 X lO" n/cm^ (£• > 0.1 MeV)

The irradiation temperatures were determined with thermal expansion vices (TED) [2] TEDs were located at the top and bottom of each subcap-

de-TABLE 3—Irradiation temperatures for

in-reactor creep specimens

Capsule

8329°

B329 B329 B334 B333 B331''

Design Temperature, °C

"Inconel 600 TED in same level as specimens

'Average peak temperatures for TEDs above and below level containing specimens

sule and were used to indicate the maximum temperature to which the imens were exposed during irradiation The results of the analysis of the TEDs are summarized in Table 3 The TED temperatures reported in Table 1 have been corrected for measured volume changes in the cladding material The nominal irradiation temperature assumes that the average irradiation temperature is the midplane coolant temperature plus 90% of the tempera-ture difference between the coolant and peak (TED) temperatures In other words, the 7-heating at the specimen locations is assumed to vary ±10% dur-ing the course of an irradiation

spec-The irradiation vehicle for the swelling specimens identified as the AAI test was of similar outer dimensions Inside were eight subcapsules, each of which contained identical specimen loadings immersed in sodium The sub-capsule temperatures were obtained by controlled gamma heat losses through an inert gas gap between the subcapsules and the capsule Design temperatures were 400,425,455,480, 510, 540, 595, and 650°C Irradiations were in Row 2 of EBR-II A low fluence experimental test of this design used TEDs to check operating temperatures, and the temperature uncertainties are estimated at ±25°C for the higher subcapsule temperatures The major factor controlling this uncertainty was found to be variations in the reactor gamma heating rate Heat transfer calculations based on those gamma heat-ing values indicate that the actual operating temperatures were lower than the design temperatures by as much as 20°C Reactor fluences given for both experiments are the product of the EBR-II flux for the appropriate reactor positions and the residence time of the vehicle in-reactor The fluence uncer-tainty is estimated to be +10%

26 EFFECTS OF RADIATION ON MATERIALS

TABLE 4—Diameter change measurements for 2'l<Cr-lMo pressurized

tube specimens contained in the AAXIV experiment [3]

13 (2.6) 28.5 (5.7)

13 (2.6) 28.5 (5.7)

13 (2.6) 28.5 (5.7) 11.5 (2.3)

27 (5.4) 11.5 (2.3)

27 (5.4) 11.5 (2.3)

27 (5.4) 11.5 (2.3)

27 (5.4)

Midwall Hoop Stress, MPa

%

0.002 0.09 0.038 0.20 0.033 0.22 0.041 0.27 0.023 0.01 0.037 0.05 0.090 0.12 0.102 0.15 -0.016 -0.06 0.308 0.69 0.483 1.85 2.290 (failed) 2.33 (failed)

TABLE 5—Swelling measurements {—t^p/pa) for I'UCr-lMo

specimens contained in the AAI test [4], po = 7.8414

Specimen Temperature, Dose, Swelling, Number °C dpa (X lO" n/cm^) %

94M6 94M7 94L6 94L7 94E6 94E7 94F6 94F7 94K6 94K7 94G6 94G7

66 77.5 76.5

99

86 120.5 83.5

116

(1.40) (1.60) (1.58) (2.07) (1.32) (L55) (1.53) (1.98) (1.72) (2.41) (1.67) (2.32)

0.12 0.28 0.08 0.14 0.09 0.08 0.10 0.05 0.09 0.01 0.17 0.22

0.2

DIAMETRAL CREEP STRAIN

(%) 0.1

FIG 1—Diameter change corrected for swelling as a function of dose at 390PCfor midwall hoop

stress levels as high as 100 MPa The solid curves define the in-reactor creep correlation prediction for the stress indicated The dashed curves define the irradiation creep contribution without swelling- enhanced creep

plotted in Figs 2 through 7 Figures 2, 3, and 4 show diameter change as a function of fluence for each of the irradiation temperatures, 390, 480, and 570°C The diametral changes shown in these figures have been corrected for volumetric changes in the specimen by subtracting the diameter change for the unstressed specimen from the stressed specimen diametral strain at each

DIAMETRAL CREEP STRAIN

FIG 3—Diameter change corrected for swelling as a function of dose at 48ff'Cfor midwall hoop

stress levels as high as 100 MPa The curves define the in-reactor creep correlation prediction with thermal creep neglected for the stress value indicated

28 EFFECTS OF RADIATION ON MATERIALS

DIAMETRAL CREEP STRAIN

l%l 1.0

FIG 4—Diameter change corrected for swelling as a function of dose at 57(fCfor midwall hoop

stress levels as high as 100 MPa Note that the 100 MPa specimen failed prior to 2.3 X / O " n/cm^ The curves define the in-reactor creep correlation prediction with thermal creep neglected

temperature The curves on each of these figures define the design tions given in the Discussion of this paper In all but one case, diameter change increases with stress (the exception being at 390°C and low fluence where the variations are within the uncertainty of the measurement tech-nique.) The high stress specimen (100 MPa) tested at 570°C had apparently failed prior to the lower fluence measurement (2.3 X 10^^ n/cm^ or 11.5 dpa)

correla-100 HOOP STRESS IMPal

FIG 5—Diameter change at 39(fC corrected for swelling as a function of hoop stress The

curves define the in-reactor creep correlation predictions

0,20

50 100 HOOP STRESS (MPa)

150

FIG 6—Diameter change at 48ff'C corrected for swelling as a function of hoop stress The

curves define the in-reactor creep correlation prediction with thermal creep neglected

HOOP STRESS (MPa)

FIG 7—Diameter change at 57CPC corrected for swelling as a function of hoop stress The

curves define the in-reactor creep correlation prediction with thermal creep neglected

30 EFFECTS OF RADIATION ON MATERIALS

at a diameter strain of 2.3% The stress dependence of the corrected diameter changes at 390, 480, and 570°C is shown in Figs 5, 6, and 7 Response at 390°C and at 480°C (Figs 5 and 6) can be adequately represented by a Hnear fit given the scatter and scarcity of data, but the onset of tertiary creep pre-vents such an analysis in Fig 7 It can also be shown by comparison of Figs

2 and 3 or 5 and 6, that in-reactor creep at the higher fluence is greater for the 380°C case than for the 460°C case Such a response can occur when ir-radiation creep is enhanced by swelling

The pressurized tube data can also provide swelling information The ameter change for the unstressed conditions can be interpreted from Eq 1 to give fractional swelling values If one assumes isotropic swelling, then swel-ling (5)

di-S = AF/Ko « 3 AD/Do (1)

where V is the specimen volume, D is the specimen diameter, and the

sub-script, 0, refers to the unirradiated condition

Therefore, swelling at 390°C to 5 X 10^^ n/cm^ or 25 dpa can be estimated

at 0.27%, whereas at 480 and 560°C, the swelling is negligible and tion of 0.18% occurs at 570°C

densifica-The swelling values from Table 5 are plotted in Fig 8 Figure 8 also shows the results of the zero stress pressurized tubes for comparison From Fig 8,

SWELLING

(%)

o DESIGN CORRELATION L— 80 120 dpa

FIG 8—Swelling results as a function of irradiation temperature Both density change and

un-stressed pressurized tube data are shown The smaller open and closed data points define the ling correlation prediction for the fluences corresponding to the swelling results shown

swel-it can be demonstrated that swelling is low in 2V4Cr-lMo to doses as high as

120 dpa Peak swelling occurs at the lowest temperature of 400°C but a

sec-ondary swelling peak is found at 540°C Based on behavior of other ferritic

alloys [4], this secondary peak is expected to arise as a result of precipitation

rather than void swelling However, the 400°C peak can be expected to be

due to void swelling A swelling rate of 0.08%/10^^ n-cm~^ or 0.016%/dpa is

predicted from the AAI results at 400°C In comparison, the pressurized

tubes produced comparable swelling values at a much lower dose This may

be an effect due to heat-to-heat, fabrication, or heat treatment variations It

may also be noted that densification is occurring at 510°C, an indication that

precipitation is continuing at high dose

Discussion

The intent of this work is to develop an understanding of the in-reactor

creep and swelling response of 2V4Cr-lMo steel and to present the results in a

format consistent for its inclusion in the Materials Handbook for Fusion

Energy Systems The latter objective requires that the results be provided in

a design equation format However, it is not within the scope of this project

to provide a complete and defendable design equation Nor is the data set

sufficient by itself to provide a clear indication of the functional dependence

required for such equations Therefore, the approach that will be taken will

be to assume the necessary functional dependence based on other martensitic

steels and experimental ferritic alloys and then to establish the values of the

necessary parameters in order to obtain an acceptable fit to the present data

sets Of considerable concern was the establishment of an in-reactor creep

equation that was compatible at high temperature with out-of-reactor

thermal creep data This required that the thermal creep dependence of

2V4Cr-lMo be obtained from the literature

Swelling Equation

The form of swelling equation used is standard and consists of two

tional relationships Void swelling (^o) is modeled using a bi-linear

func-tional relationship with three adjustable parameters: a swelling rate, R; a

swelling incubation parameter, r; and a transition parameter, a Each of

these parameters can be specified as a function of temperature [5]

Concur-rently, densification (D) is modeled using a functional relationship with two

adjustable parameters: a steady-state density, D*, and a transition

parame-ter, X, and again each of these parameters can be specified as a function of

temperature The equations are as follows

Swelling = - — (%) = So-D (2)

Vo