Astm stp 1046 v1 1990

These papers move beyond simple predictions of the dose or temperature dependence of void swelling to explore more fundamental defect behavior and the complex interactions between the va

STP 1046

Effects of Radiation on

Materials: 14th International Symposium (Volume I)

N H Packan, R E Stoller, and A S Kumar, editors

1916 Race Street

Philadelphia, PA 19103

Library of Congress Cataloging-in-Publication Data

Effects of radiation on materials: 14th international symposium/

N H Packan, R E Stoller, and A S Kumar, editors

(STP; 1046)

Papers from the Fourteenth International Symposium on Effects of Radiation on Materials, held June 27-30, 1988 in Andover, Mass and sponsored by ASTM Committee E-10 on Nuclear Technology and

Applications

Includes bibliographies and index

ISBN 0-8031-1266-1

1 Materials Effect of radiation on Congresses I Packan, N

H (Nicholas H.) II Stoller, R E (Roger E.), 1951-

III Kumar, A S (Arvind S.) IV International Symposium on

Effects of Radiation on Materials (14th: 1988: Andover Mass.)

V ASTM Committee E-10 on Nuclear Technology and Applications

VI Series: A S T M special technical publication: 1046

Peer Review Policy

Each paper published in this volume was evaluated by three peer reviewers The authors addressed all of the reviewers" comments to the satisfaction of both the technical editor(s) and the A S T M Committee on Publications

The quality of the papers in this publication reflects not only the obvious efforts of the authors and the technical editor(s), but also the work of these peer reviewers The ASTM Committee on Publications acknowledges with appreciation their dedication and contribution

of time and effort on behalf of ASTM

Printed in Baltimore MD April 1990

Foreword

Effects of Radiation on Materials: Fourteenth International Symposium was presented at Andover, MA, 27-30 June 1988 The symposium was sponsored by ASTM Committee E- l0 on Nuclear Technology and Applications N H Packan Oak Ridge National Laboratory, presided as chairman of the symposium with R E Stoller, Oak Ridge National Laboratory, and A S Kumar, University of Missouri-Rolla, as vice-chairmen There are two resulting

Special Technical Publications (STPs) from the symposium: Effects o f Radiation on Materials:

Fourteenth International Symposium (Volumes I and 11), S T P 1046 and Reduced-Activation Materials f o r Fusion Reactors, S T P 1047

Contents

Overview

MICROSTRUCTURES: FERRITICS

Behavior and Microstrnctere of Ferritic Steels Irradiated in the Phenix Reactor

DIDIER GILBON, JEAN-LOUIS SI~RAN, RICHARD CAUVIN, ANTOINE FISSOLO,

ANA ALAMO, FRAN(~OIS LE NAOUR, AND VIVIANE LI~VY

Void Formation and Helium Effects in 9Cr-IMoVNb and 12Cr-lMoVW Steels

Irradiated in HFIR and FFTF at 4 0 0 ~ J MAZIASZ AND

RONALD L KLUEH

Microstructural Change in Ferritic Steels Under Heavy ion Irradiation

KOICHIRO HIDE, NAOTO SEKIMURA, KOJI FUKUYA, HIDEO KUSANAGI,

MASAFUMI TAGUCHI, TATSUO SATAKE, YOSHIO ARAI, MASASHI IIMURA,

HIROSHI TAKAKU, AND SHIOR! ISHINO

Neutron Irradiation Damage in Ferritic Fe-Cr AlloyS DAVID S GELLES

Influence of Structure and Phase Composition on ICr13Mo2NbVB Steel

Mechanical Properties in Initial, Aged, and Irradiated Statesm

V S AGUEEV, V N BYKOV, A M DVORYASHIN, V N GOLOVANOV,

E A MEDVEDEVA, V V ROMANEEV, V K SHAMARDIN, AND

A N VOROBIEV

Effects of Thermal Aging on Precipitate Development of Alloy H T 9 - -

PO-SHOU CHEN AND ROY C WILCOX

Swelling Suppression in Phosphorus-Modified Fe-Cr-Ni Alloys During Neutron

lrradiation EAL H LEE AND NICOLAS H PACKAN

Microstructural Evolution of Neutron Irradiated Fe-Cr-Ni Alloys at 495"C

in Response to Changes in the Heliam/DPA R a t i o - - J F STUBBINS,

J E NEVLING, F A GARNER, AND R L SIMONS

133

147

Microstructurai Development of Titanium-Modified Austenitic Stainless Steel

Under Neutron Irradiation in H F I R up to 57 dpa MASAHIDE SL ZtKI,

S H O Z O H A M A D A , PHILIP J M A Z I A S Z , M1TSUO P T A N A K A , AND

A K I M I C H I H I S H I N U M A

The Microstructurai Evolution and Swelling Behavior of Type 316 Stainless

Steel Irradiated in H F I R - - s H o z o HAMADA, MASAHIDE SUZUKI,

Ion Bombardment Radiation Damage Studies of Fusion-Relevant Austenitic and

Ferritic A l l o y s - - D A V I D J M A Z E Y , SUSAN M M U R P H Y , G PETER W A L T E R S

High Temperature Embrittlemeut of Hastelloy X After Low Fluence Neutron

I r r a d i a t i o n - A n Effect o f Helium? BERNHARD A THIELE,

H E R B E R T S C H R O E D E R , WILTO K E S T E R N I C H , AND FLORIAN S C H U B E R T 271

On the Role of Helium in High Temperature Embrittlement of Irradiated

I M N E K L Y U D O V , L S O Z H I G O V , AND A A P A R K H O M E N K O 295

Helium-Induced Degradation in the Weldability of an Austenitie Stainless Steel -

H U A T LIN, STEVE H G O O D S , MARTIN L GROSSBECK, AND BRIAN A CHIN 301

lsotopically Alloyed Injector Foils for Helium Effects Research in Mixed-Spectrum

R e a c t o r s - - L O U i S K MANSUR AND W I L L I A M A C O G H L A N 315

The Growth of Bubbles in Pure Aluminum During and After Irradiation with

6 0 0 M e V P r o t o n S - - F R A N C O I S P A S C H O U D , R G O ' I - F H A R D T , D G A V I L L E T ,

Cavitation and Embrittlement in Tritium Exposed Copper STEVEN n GOODS 340

Modeling Studies on the Precipitation of Krypton After Implantation into Metals - JEFFREY REST, R O B E R T C B I R T C H E R , AND A N S H E N G S LIU 353

Simulation of Solid Inert Gas Bubbles in Metals -SUBRAMANYAM SRINIVASAN

A S H O K K T Y A G I , AND K A N W A R KRISHAN 364

R A D I A T I O N - I N D U C E D S E G R E G A T I O N OR PHASE C H A N G E S

Radiation-Induced Segregation of Phosphorus in Nickel and Fe-Cr-Ni Alloys -

JONATHAN M PERKS, COLIN A E N G L I S H , AND MIKE L JENKINS 379

Characteristics of Radiation-Induced Solute Segregation in Candidate and Model

Ferritic A I I o y s m T A K E O M U R O G A , ATSUSHI Y A M A G U C H I , A N D

Segregation to Surfaces in Irradiated Stainless Steels -EDWARD P S I M O N E N ,

Irradiation-Assisted Stress Corrosion Cracking and Grain Boundary Segregation

in Heat Treated Type 304 SS -ALVlN J JACOBS, ROBERT E CLAUSING

LEE H E A T H E R L Y , A N D R I C H A R D M K R U G E R 424

Irradiation-Induced Solid Solution Decomposition Enhances Point Defect

R e c o m h i n a t i o n - - v v BRYK, V N V O Y E V O D I N , V F Z E L E N S K Y , 1 M

Microstructural Development in Neutron Irradiated ZircaIoy-4 -WALTER J S YANG 442

Electron Irradiation-Induced Amorphization of Precipitates in Zircaloy 2 - -

A R T H U R T M O T T A , D O N A L D R O L A N D E R , A L B E R T J M A C H I E L S 457

M I C R O S T R U C T U R A L M O D E L I N G

The Influence of Microstructure and Solutes on Void Formation and Void Growth

in Irradiated MaterialS ROGER E S T O L L E R AND LOUIS K MANSUR 473

The Stochastic Instability of a Spatially Homogeneous Void Distribution in

Irradiated Metals: The Distribution Function A p p r o a c h - -

The Slowing Down of Dislocation Climb by Voids at Later Stages of Material

Influence of Small Bubble Sink Strength on the Swelling Rate of Voids

in Austenitic Stainless Steel AIS! 316 L -JOHANNES TENBRINK

Nucleation of Helium Bubbles at Grain Boundaries During Irradiation

Void Ensemble Response to Periodical Variations of Irradiation C o n d i t i o n s - -

F U N D A M E N T A L DEFECT B E H A V I O R

A n HVEM Study of Displacement Cascade Damage in Nb~Sn at 13 K MARQUIS

A KIRK, M I C H A E L C B A K E R , B E R N A R D J KESTEL, H A R A L D W W E B E R , AND

A Study of Cascade Cluster Formation and Interaction in Ion-Irradiated Gold by

The Production Rate of Freely Migrating Defects in Self-Ion Irradiated Nickel and

Alternative Polarity Recombination Centers of Point DefectS -

Helium Bubbles in Copper Studied by Positron Annihilation -G AMARENDRA,

B VISWANATHAN, AND K P GOPINATHAN

Study of Helium Diffusion to Grain Boundaries by the Method of

STP1046-EB/Apr 1990

Overview

The Fourteenth International Symposium on Effects of Radiation on Materials was held

on 27-30 June 1988 in Andover, MA This biennial symposium series commenced in 1956 and has served as a major international forum for the exchange and discussion of both the fundamental and technological aspects of behavioral changes in materials exposed to radia- tion environments The high level of participation at the latest symposium required four full days of conference sessions, and the peer-reviewed proceedings are being published in three volumes

The papers from the first three days of the symposium appear in the two volumes of this ASTM Special Technical Publication (STP) 1046 Volume [ encompasses radiation damage- induced microstructures; point defect, solute, and gas atom effects; atomic-level measure- ment techniques; and applications of theory Volume II includes mechanical behavior, all papers dealing with pressure-vessel steels, breeder reactor components, dosimetry, and nuclear fuels The fourth day of the symposium was devoted to the single topic of reduced- activation materials, including austenitic, ferritic, and vanadium alloys, for future fusion reactors; these papers are being published in a companion volume: ASTM STP 1047, Re- duced-Activation Materials for Fusion Reactors

The first two sections of Volume I, Microstructures: Ferritics and Microstructures: Aus- tenitics, deal with the effects of radiation on the structures of alloys being developed for

high-dose environments, such as the first wall of a fusion reactor One notable theme is the profound influence of certain - m i n o r " (that is, present in low concentration) but critical alloying elements, such as titanium, phosphorous, and others, primarily through the unique phases and microstructures they induce Examples are offered of fine dispersions of pre- cipitates acting as distributed point defect and gas atom traps with the result that the growth

of large voids or bubbles is inhibited Both neutron and charged-particle irradiations have been employed in these papers, while analytical transmission electron microscopy (TEM) seems to be the major technique of investigation

In Gas Effects, a large number of papers (12) explore various aspects of what is probably

the most pernicious nuclear transmutation product, helium If it is not hindered from col- lecting in bubbles at grain boundaries, the frequent result is helium embrittlement, which becomes a principal constraining factor for operation at high irradiation temperatures Among the papers presented are studies of helium embrittlement in ferritic alloys, in pure nickel and nickel-base alloys, a direct comparison between austenitic and martensitic steels, helium effects in the different regions of stainless steel welds, and bubble formation in pure copper and aluminum There are also fundamental studies of bubble formation for various inert gas species, and a technique paper offering a new way to introduce helium in mixed- spectrum reactor experiments by the use of nickel-bearing foils adjacent to the specimens The section on Radiation-Induced Segregation or Phase Changes is an important one

2 EFFECTS OF RADIATION ON MATERIALS

because such radiation-induced modifications threaten to undo some of the hard-won gains

in resistance to radiation damage obtained from increasingly sophisticated alloying An alloy matrix can be depleted of critical constituents, and if segregation or precipitation occurs at the grain boundaries, the material can be embrittled These papers explore the extent of segregation of such elements as phosphorous, silicon, and nickel in mainly austenitic or ferritic alloys, typically with the aid of surface analytical tools, such as Auger electron spectroscopy or secondary ion mass spectrometry (SIMS) Included are two papers dealing with the amorphization under irradiation of precipitates in zirconium alloys

to model radiation effects has reached a high level of sophistication These papers move beyond simple predictions of the dose or temperature dependence of void swelling to explore more fundamental defect behavior and the complex interactions between the various ex- tended defects that are observed in irradiated materials Topics include: the direct influence

of sinks, such as dislocations, precipitates and bubbles, on void swelling: dislocation loop formation; the formation of void and loop lattices: and possible void-dislocation interactions that could lead to swelling saturation Two papers that are of specific interest to the fusion materials community consider the influence of pulsed irradiation and the important problem

of helium bubble formation at grain boundaries

of radiation damage The techniques used in the work reported here include internal friction, SIMS, and heavy-ion irradiation in a high-voltage electron microscope The materials ranged from austenitic alloys to the A 15 superconductors These results provide critical information about defect behavior and parameters that are required for modeling studies The papers discuss the nature of the displacement cascade, the small defect clusters that are formed as the cascade collapses, and the fraction of the originally-created defects that survive and diffuse out of the cascade, It is primarily these latter defects that lead to the observable effects of radiation The fate of these freely migrating defects and the helium that is produced

by nuclear transmutation reactions is also discussed

specialized applications using effects from the domain of physics as novel tools for the probing

of atomic-level processes The techniques employed are positron annihilation, internal fric- tion, the Mossbauer effect, and nuclear gamma resonance

Nicolas H Packan

Oak Ridge National Laboratory, Oak Ridge,

TN 37831: symposium chairman and coed- itor

Oak Ridge National Laboratory, Oak Ridge,

TN 37831; symposium vice-chairman and coeditor

Arvind S Kumar

University of Missouri-Rolla, Rolla, MO 65401: symposium vice-chairman and coeditor

Microstructures: Ferritics

Didier Gilbon, ~ Jean-Louis Sdran, ~ Richard Cauvin,

A n t o i n e Fissolo, ~ A n a A l a m o , ~ Francois L e Naour, ~ Viviane L ~ v y

Behavior and Microstructure of Ferritic

Steels Irradiated in the Ph6nix Reactor

and L~vy, V., "Behavior a d Microstmctm~ of Fenilk Steels Irradiated in the Ph,~nix Re-

STP 1046, N H Packan, R E Stoller, and A S Kumar, Eds., American Society for Testing and Materials, Philadelphia, 1989, pp 5-34

ABSTRACT: This paper deals with the irradiation behavior of three ferritic steels, namely FI7 (17Cr), EMI2 (9Cr-2MoNbV) and EMI0 (9Cr-lMo) These alloys were irradiated up to

100 dpa in Ph~nix as samples or wrapper tubes The immersion density measurements confirm their high swelling resistance, but the tensile and impact tests reveal great differences in mechanical properties The ductile-brittle transition temperature (DBTT) of Fi7 is strongly increased, and compared to an aging treatment, the irradiation amplifies and shifts the em- brittlement towards lower temperatures In contrast to FI7, the mechanical properties of EM10 are unaffected by irradiation, while EMI2 has an intermediate behavior

The transmission electron microscopy (TEM) examinations show that all the small density changes come from irradiation-induced voids and that the embrittlement of F17 results from a' phase formation enhanced by irradiation In conclusion, EMI0 is by far the most attractive candidate for wrapper applications Its fully martensitic structure provides an improved swelling resistance and its chemical composition should inhibit the microstructural instabilities that are responsible for ~he embrittlement of F17 and EMI2

ing, mechanical properties, embrittlement, microstructure

For many years, austenitic stainless steels have been widely used as wrapper and fuel pin cladding materials in fast breeder reactors In spite of continuous progress, their swelling resistance does not seem adequate to reach the target doses necessary for economical com- petitive reactors of the future O n the other hand, ferritic steels are becoming more attractive candidates for in-core structural applications Much better swelling resistance has been demonstrated by a great number of experiments and many ferritic steels have sufficient strengths in out-of-pile conditions However, the effect of irradiation on their mechanical properties still requires investigation, with a particular emphasis on both creep behavior at high temperatures for cladding applications, and change in ductile-brittle transition tem- perature (DBTI') for wrapper applications

Several types of ferritic steels ate under development for fuel elements in the French fast breeder reactor program This paper deals with the irradiation results of three of these alloys, namely F17 (17 Cr), EM12 (9 CR-2 MoNbV), and EM10 (9 Cr-1 Mo) First we describe the swelling behavior of these three alloys, then show the effect of irradiation on Research scientists, CEA/IRDI/DMECN/D TECH/SRMA, Cen-Saclay 91191, GIF-sur-YVETTE C6dex, France

6 EFFECTS OF RADIATION ON MATERIALS

TABLE 1 Chemical composition of the three alloys (wt % )

The two o t h e r alloys are used in a normalized and t e m p e r e d state (Table 2) T h e E M I 2 steel (9 Cr-2 M o N b V ) has a duplex structure c o m p o s e d of about 30% ferrite and 70% martensite T h e grain and lath boundaries are d e c o r a t e d mainly by precipitation of M_,3C, carbides with a few (Nb, V ) C particles The unstabilized E M I 0 steel (9 Cr-1 Mo) is a fully martensitic alloy Its prior-austenite grain boundaries and some of its lath boundaries are also d e c o r a t e d with M23C~ carbides and its dislocation density is lower than that of E M 1 2 (Fig 1)

Irradiation Conditions

The F17 steel was irradiated in Ph6nix as w r a p p e r tube in two fissile subassemblies

W r a p p e r B1 was irradiated for 9552 h in the central part of the core up to a dose of 62.5 dpa, while w r a p p e r B2 was irradiated for 19 845 h in the peripheral part of the core up to

TABLE 2 Heat treatments and structure of the three alloys

" c.w means cold-worked

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE 7

FIG l Microstructure of the unirradiated FI7, EM12, and E.'~IIO steels

8 EFFECTS OF RADIATION ON MATERIALS

a dose of about 100 dpa After irradiation, specimens were cut from these two wrapper

tubes by spark machining techniques in order to perform density measurements and tensile

and Charpy tests

The EM10 and EM12 steels were irradiated as tensile and Charpy test specimens in an

experimental rig to a dose of up to 40 dpa These specimens were cut from unirradiated

wrapper tubes or sheets These two fabrication routes give rather similar results and will

not be further distinguished Lastly, the EM12 alloy was also irradiated as empty cladding

tubes in another experimental rig and their swelling values were repeatedly measured at

intermediate doses during irradiation up to about 100 dpa

In this paper, all the doses are expressed in dpa NRT and the temperatures are calculated

values along the length of the fissile column In the case of wrapper tubes, the orientation

of each of the six faces with regard to the center of the core is taken into account

Mechanical Tests

The impact tests were carried out using quasi-standard Charpy V-notch specimens, ttow-

ever, the thickness (3.5 mm) was required as a consequence of the thickness of the wrapper

tube, and the available energy of the instrumented Tinius-Olsen pendulum had to be reduced

to 120 J to get an impact velocity of about 3 m/s

The tensile tests were conducted at room temperature, at 180~ and at irradiation tem-

perature using a strain rate of 3 x 10 ~s-~ For the F17 steel, flat tensile specimens were

cut from the irradiated wrapper tube, while for the EM10 and EM12 alloys, cylindrical

tensile specimens (2 mm in diameter) with screwed heads were prepared prior to irradiation

For the high temperature tests, a heating rate of 2~ was applied and a stabilizing time

of 0.25 h was necessary

For the impact as well as the tensile tests, specimens were cut along transversal and

longitudinal directions of the wrapper tubes After irradiation, for these three ferritic steels

and for the dose range investigated in this paper, these two sampling conditions only induce

slight differences and will therefore not be distinguished

Transmission Electron Microscopy

All the TEM examinations were performed on a Philips EM 430 microscope equipped

with a LaB, filament and operated at 300 kV This high voltage limits the image distortion

due to the magnetic properties of these ferritic steels The TEM samples were taken from

the undeformed parts of the Charpy or tensile specimens tested at the lowest temperatures

To limit uncertainties on irradiation dose and temperature, transversal specimens were

generally used After mechanical thinning down to 100 lxm, thin foils with a diameter of 3

mm were prepared by electropolishing in a twin jet device

The swelling values were calculated from the averaged void volume and number density

based on foil thickness measured by stereoscopy The microstructural evolution (dislocation

loops and lines, precipitation, and so on) has been studied by the usual techniques of

microdiffraction and bright and dark field imaging The scanning and energy dispersive

devices ( S T E M / E D S ) also allowed X-ray microanalysis on some aged and a few irradiated

samples For these experiments, an electron probe diameter of about 7 nm was used

Results

Swelling Behavior

F17 Wrapper Tubes The values of dose and swelling have been plotted versus irradiation

temperature for the two F17 w~Zapper tubes in Fig 2 As the irradiation of Wrapper Tube

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE 9

FIG 2 Evolution of dose and swelling of the FI7 steel versus irradiation temperature (1: Wrapper

BI; 2 and 5: Wrapper B2 faces 2 and 5)

B2 was done in the peripheral part of the core, two different sets of curves have been

obtained from two opposite faces (2 and 5) of the wrapper These results demonstrate the

high swelling resistance of the ferritic steels, since the swelling values never exceed 0.7%

for doses up to 100 dpa In the lower temperature range, the swelling increases quickly and

reaches its maximum value at about 420~ Above 440~ the swelling sharply decreases

while the dose is still increasing, and finally becomes negligible above 470~ ( - 0 1 % )

The evolution of swelling with irradiation dose is analyzed in Fig 3 The small number

of data available leads to a rather wide temperature sampling, but in spite of the scatter,

10 EFFECTS OF RADIATION ON MATERIALS

l l l l 1

r,., 4,1

i t )

r , , ~ , r - II

GILBON ET AL ON BEHAVIOR A N D M I C R O S T R U C T U R E 11

+I

C,O

O) I'- ~I"

II I.I I,-

1 2 EFFECTS OF RADIATION ON MATERIALS

GILBON ET AL ON BEHAVIOR AND M I C R O S T R U C T U R E 1 3

14 EFFECTS OF RADIATION ON MATERIALS

the swelling seems to increase linearly with irradiation dose and the highest swelling rate is

observed for temperatures below 430~ (1 x 10 -'~ However, this swelling rate,

which is much lower than for any austenitic steel, still might not be the steady state value

Moreover, Fig 2 shows that the swelling of the F17 steel also depends on the dose rate

Thus, when the doses are similar up to 450~ the swelling values for B1 (18 x 10 -7

dpa/s) are higher than for Face 5 of B2 (8 x 10 7 dpa/s), and these values are similar to

those of Face 2 of B2 (14 x 10 -~ dpa/s) in spite of the difference in dose

EM 10 and EM12 Samples For the EM10 and EM12 specimens irradiated at up to 40

dpa in an experimental rig, the density measurements never showed swelling values higher

than 0.1% over the temperature range investigated However, the EM12 steel has been

irradiated to higher doses (100 dpa) as empty tubes This allows comparison of swelling

behavior between EM12 and F17 for temperatures close to 400~ (Fig 3) The duplex EM12

alloy is also highly swelling resistant and significant values are onlv observed for low irra-

diation temperatures At about 400~ the swelling rate of E M I 2 might be a little higher

than for F17, but its incubation dose also seems larger

Tensile Tests

F17 Steel The results of tensile tests performed at room temperature on Wrapper Tube

B1 (60 dpa, 9552 h) are compiled in Figs 4a and b Compared with the unirradiated material,

all the specimens irradiated below 540~ show an increase in strength and a reduction of

elongation, with this hardening effect becoming more important at lower irradiation tem-

perature However, an aging treatment of 10 000 h also leads to significant hardening of

the F17 steel, so that the main effect of irradiation is to shift the maximum in tensile strength

and the minimum in elongation by 50 to 100~ towards lower temperatures and to reduce

the total elongation to less than 2% at about 400~ On the other hand, the tensile tests

performed at 180~ show that the two wrapper tubes have similar behaviors, in spite of

large differences in their irradiation time, fluence, and dose rate (Figs 4c and d)

Finally, the various values of strength and elongation measured at the irradiation tem-

perature are plotted in Fig 5 Compared with the unirradiated material, these results confirm

the important hardening of the wrapper tubes for temperatures below 500~ and show a

weak tendency towards softening for higher temperatures The general agreement between

the tensile tests on the two wrapper tubes can again be observed, and the comparison with

the tensile tests performed on samples aged for only 2 000 h further confirms the shifts of

strength and ductility towards lower temperatures

To summarize, a clear hardening effect of neutron irradiation has been demonstrated for

temperatures below 420~ in this narrow range where the dose rate quickly increases, all

the tensile tests indicate significant reductions of elongation and increases in strength This

contribution of irradiation to hardening is described in Fig 6, which shows that saturation

values are reached after a dose of about 40 dpa

EMIO Steel Figure 7 allows comparison of tensile properties of the EM10 steel irradiated

at a maximum dose of 40 dpa with the F17 wrapper tube irradiated at 60 dpa These results

show that the behavior of EM10 is much less affected by irradiation temperature Thus, for

temperatures above 500~ the total elongation of EM10 remains below 20%, while the

elongation of F17 may exceed 80% However, the most striking feature is that in contrast

to F17, the values of strength and ductility of irradiated EM10 are very close to those

measured on unirradiated samples In the dose range that has been studied (<40 dpa), only

a small amount of hardening can be observed at temperatures below 500~ Similar data

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE 15

on EMI2 are not yet available and the comparison between the three alloys can only be

made with the impact tests

Impact Tests

F17 Wrapper Tubes The evolution of the total absorbed energy during Charpy tests is

plotted with test temperature for the unirradiated F17 steel and for specimens from the

hexagonal Tube B1 (60 dpa) in Fig 8 As already has been described in an earlier paper

[1], the lower the irradiation temperature, the higher the DBTT and the lower the upper

and the upper shelf energy (USE) falls to 70J/cm z, instead of - 5 0 ~ and 200 J/cm z re-

spectively, for the unirradiated material

Figure 9 shows the results of Charpy tests (DBTT and USE) performed on the two F17

wrapper tubes As was found for the tensile tests, the results for B1 (60 dpa high dose rate)

and for B2 (100 dpa, low dose rate) are in perfect agreement The evolutions of DBTT and

USE versus irradiation temperature further establish that the embrittlement of the irradiated

F17 steel is all the more prominent as the irradiation temperature is reduced, but even above

540~ the value of the D B T T remains 100~ higher than for unirradiated material

Comparison of data on irradiated material with the impact tests performed on Charpy

specimens aged for 10 000 h again shows that irradiation shifts embrittlement towards lower

temperatures and amplifies the degradation of mechanical properties of the F17 alloy

Finally, the values of D B T F and USE are plotted versus irradiation dose for irradiation

10) These two curves show that in contrast to the tensile properties, shifts in D B T F and

reductions in USE occur at very low fluence and that the embrittlement does not progress

with increasing dose

EMIO and EMI2 Steels Figure 11 allows comparison of the impact behavior for EMI0

and EM12 samples irradiated to 40 dpa and F17 steel irradiated to 60 dpa In contrast to

F17, and as in tensile tests, the properties of the EM10 alloy are nearly unchanged by

irradiation Slight DBTT shifts and USE reductions are only observed at the lower and

The EM12 steel does not become as brittle as the F17 alloy, since its DBTT values remain

below 100~ over the whole range of irradiation temperature As expected, for temperatures

are similar for EM12 and F17 and the USE of the EM12 steel becomes much lower

Microstructural Evolution

Table 3 tabulates the irradiation conditions and the results of density measurements for

all the samples that have been studied by TEM These T E M examinations had to obey two

main objectives First was to check whether or not the very small density changes observed

for the irradiated F17 and EM12 steels came from classical mechanisms of irradiation-induced

void nucleation and growth, as well as to get more information on the swelling behavior of

the ferritic steels Second was to find and correlate the microstructural behavior with the

evolution of the mechanical properties, as highlighted by tensile and impact tests In the

case of the F17 steel, the microstructure of aged samples was also studied to help to discern

the role of nonthermal process in the evolution during irradiation

16 EFFECTS OF RADIATION ON MATERIALS

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE 17

FIG 6 F17 tensile tests at irradiation temperature Effect of dose on strength and elongation for

irradiation temperatures <42WC {unirradiated values indicated by f )

Void Microstructure

F17 Steel Figure 12 shows that over a wide range of temperatures, the fast neutron

irradiation produced voids in the F17 steel For the samples irradiated at low temperature

Above 450~ the void density decreases sharply and their spatial distribution is no longer

uniform Most of the voids are associated with elongated precipitate particles and some

alignments of voids and precipitates are often observed However, at about 500~ in addition

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE 19

to a low density of large voids, a high density of very, small cubic voids may also be observed

(Fig 12) Finally, at 538~ large voids are very rarely observed in association with carbides

present prior to irradiation, but a few very tiny voids can again be detected in small areas

within the grains

All the swelling data as measured by TEM are given in Table 4 The various swelling

parameters follow the usual evolution trends: void size increases and void density decreases

with increasing temperature Moreover, the TEM values are in agreement with the im-

mersion density results and a significant void swelling is only observed in the lower tem-

perature range However, the occurrence of an additional population of small voids at 500~

and above is worth noticing Irradiation to higher doses is necessary to see whether or not

the growth of these small voids can lead to significant swelling values at temperatures above

500~

EM12 and EMIO Steels Irradiation-induced voids were also observed in the EM12 empty

tubes and Fig 13 allows comparison of the microstructure produced in the ferrite and

martensite regions of this duplex allov At 413~ the local values of void size, density, and

swelling (Table 5) are quite similar in ferrite and martensite However, void-denuded zones

are often observed near the lath boundaries and the average swelling is lower in the mar-

tensite than in the ferrite phases This difference is much more pronounced at about 440~

since the void distribution is still uniform in ferrite grains, whereas voids are only observed

in a few martensitic laths At 460~ the irradiation induces no more swelling, neither in the

ferrite, nor in the martensite

The three TEM examinations performed on the irradiated EM10 samples did not reveal

any voids However in this low dose range (<40 dpa), similar examinations were also

unsuccessful for EM12 Further irradiations to higher doses are still necessary to conclude

that the fully martensitic EM10 alloy is more swelling resistant

Precipitation

FI 7 Steel Before examining irradiated material, the precipitate development was studied

for samples aged at 400 and 550~ ( - 1 0 000 h) and at 450 and 500~ ( - 2 5 000 h) In the

material aged at 450 and 500~ a high density of tiny particles is revealed under weak strain-

contrast conditions (Fig 14, left) A t 400~ these particles are within resolution limits and

were not detected at 550~ Diffraction spots originating from these particles were not

observed, but all their microstructural features are similar to those described by many authors

[2-6] for the a ' phase Moreover, X-ray microanalysis experiments were performed in the

thinnest parts of the sample aged at 500~ The results, which include a small amount of

matrix contribution, confirmed that the observed particles are rich in chromium and silicon

(Table 6)

In the irradiated samples, larger a ' phase particles were observed over the whole range

of temperature investigated (400 to 540~ In contrast to the aged material, these particles

are easily observed at the lower irradiation temperatures (Fig 14, right) Their distribution

is quite uniform up to 460~ but then becomes patchy at temperatures above 500~ Thus

at 540~ wide a'-free zones can be observed along grain or subgrain boundaries, disloca-

tions, and stringers of precipitates The quantitative results concerning c,' formation in the

irradiated F17 steel are shown in Table 4 The evolutions of the a ' density and size with

temperature are quite similar to those previously described for the irradiation-induced voids

Moreover at 500~ and above, the additional population of small voids and the a ' particles

are observed in the same regions

In the sample aged at 550~ a ' phase particles were not detected but additional large

20 EFFECTS OF RADIATION ON MATERIALS

I / /

I / / /

22 EFFECTS OF RADIATION ON MATERIALS

FIG 8 F17 Impact tests Evolution of the total absorbed energy versus test temperature for different

irradiation temperatures (Wrapper B1)

precipitates were found along the grain boundaries (Fig 14) Microdiffraction patterns and

X-ray microanalysis identified them as sigma phase precipitates (Table 6) A few globular

sigma phase particles were also observed in the samples irradiated at 540~ Moreover,

above 460~ more and more elongated particles are present as the irradiation temperature

increases These particles are often associated with voids Microdiffraction patterns also

suggest that they are sigma phase but further experiments and X-ray microanalysis data are

still necessary

EM12 and EMIO Steels In the EM12 cladding piece irradiated at 460~ a high density

of precipitates identified as chi phase (Table 6) was produced both in ferrite and martensite

(Fig 13) Wide denuded zones were observed along grain and subgrain boundaries and the

dislocation lines were generally connected to these particles A t 440~ similar differences

in void and precipitate distributions were found between the ferritic and martensitic regions:

the chi phase particles were uniformly distributed in ferrite grains while they were only

present within a few of the martensitic laths Finally, at the two lowest irradiation temper-

atures, a high density of small unidentified particles were found decorating the dislocation

network

In the three EM10 samples that have been studied, the precipitation present prior to

irradiation seems to be unaffected and no new phases have appeared By contrast, the same

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE 23

FIG 9 F17 impact tests Effect of irradiation (l and 2: Wrappers BI and B2) or aging (A: aged

10 000 h) temperature on D B T T and USE

In the F17 steel irradiated at temperatures close to the peak swelling one, the microstruc-

ture is composed of a high density (pa ~ 5 • 10 t~ cm ~) of tangled dislocations (Fig 15)

At intermediate temperature where the swelling is already negligible, the dislocation lines

24 EFFECTS OF RADIATION ON MATERIALS

become straighter and their density declines rapidly with increasing temperature, but a high

number of small loops can be observed between the dislocation segments Finally, at 540~

the dislocation density is quite similar to that observed in unirradiated samples, and only a

few loops can be detected in the regions where the a ' phase is present

Figure 15 also allows observation of the evolution of the dislocation microstructure with

temperature in the EM10 samples irradiated at low dose (<40 dpa) At low temperature

(T <- 450~ the dislocation density is increased and a high number of small loops are induced

FIG I t - - I m p a c t tests" on the ['17 (F), E M 1 2 (D), and E M I O (M) steels Evolution o f dose, D B T T ,

and USE with irradiation temperature (f, d , m : unirradiated)

26 EFFECTS OF RADIATION ON MATERIALS

TABLE 3 -Irradiation characteristics and swelling of the samples studied

a V / V (%)

lath structure in the irradiated material seem similar to those observed prior to irradiation

Discussion

Swelling Behavior and Microstructural Evolution

This study has shown that our three ferritic alloys, which widely differ in chemical com- position and structure, are highly swelling resistant The F17 and EM12 steels irradiated up

to 100 dpa N R T exhibit only small density variations ( < 1 % ) , with swelling peak at about

FIG 12 Effect Qf" temperature on the void microstrz<cture in the F17 ~tee/

GILBON ET AL, ON BEHAVIOR AND MtCROSTRUCTURE

Additional population of small voids

small changes in density result from the nucleation and growth of irradiation-induced voids

These general trends are in agreement with published data on the neutron irradiation effects

in simple as well as commercial ferritic steels The origin of intrinsically low swelling claimed

information to these discussions about general mechanisms However our irradiation results

and TEM examinations do allow us to discuss the effects of chemical composition and initial

structure on the swelling behavior and the associated microstructural evolution of each one

of the three ferritic steels and to compare these data with the literature

In the fully ferritic F17 steel, a uniform distribution of voids was induced by irradiation

American Iron and Steel Institute (AISI) 430 F (UNS543020) are quite similar, Gelles

[5,6,12] only observed very few voids in this alloy irradiated at high doses (125 dpa)

However+ the AISI 430 F was solution treated (1066~ lh water-quenched (W.Q.)) while

in our F17 steel, a recrystallization treatment (cold-worked (c.w.) 20% + 800~ 15 min)

induced a heavy precipitation of M,~C~ carbides The difference in swelling response could

result from the difference of carbon content in the matrix, since this element is known

to reduce the swelling in the ferritics and partly explains the difference between simple

The second important feature of the microstructural evolution of the aged or irradiated

F17 steel is the ~' phase formation The decomposition of the iron-chromium solid solution

and the precipitation of the a ' phase at aging temperatures below 516~ is a well known

has already been observed in irradiated simple alloys (Fe-Cr Fe-Cr-Mo [3-5]) as well as

c~' phase was easily observed over the whole range of irradiation temperature For irradiation

cations, and subgrain boundaries suggests the existence of weak binding interactions between

vacancies and chromium atoms in solution These denuded zones arise because the pref-

erential exchange of chromium with vacancies migrating towards the sinks leads to a cor-

responding flow of solute atoms in the opposite direction This interaction, which enhances

point defect recombination, has already been proposd by Little and Stow [3] to explain the

28 EFFECTS OF RADIATION ON MATERIALS

FIG 13 Effect of ternperature on the void and precipitate microstructure in ~he E ~412 cladding material

(M: rnartensite; F: ferrite)

beneficial effect of a 1 to 5% chromium addition on the swelling of iron, while solute

segregation away from the sinks partly explains the progressive increase in swelling for

higher chromium contents [3-4] In F17, the correlation between the swelling maximum

and the high density of a ' particles at low temperature is worth noticing Thus, for higher

irradiation doses, the subsequent growth of void population will take place in a chromium-

depleted matrix, and this could lead to progressive increase in swelling rate [3]

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE

TABLE 5 -Results of TEM examinations on the irradiated EMI2 steel

29

In spite of significant differences in chemical composition and initial structure, irradiation

leads to rather similar swelling results in F17 (17 Cr) and EM12 (9 Cr-2 MoNbV) (Tables

4 and 5) However, in this duplex alloy, a difference in average swelling was found between

the ferrite and martensite regions Similar differences have already been observed by several

authors [4-6,11] Since at 413~ the local values of mean void size and density inside the

martensitic laths are the same as those measured in ferrite, this difference in average swelling

has mainly to be attributed to the lath boundaries acting as efficient sinks for point defect

elimination This mechanism could also explain the enhancement of this swelling difference

between ferrite and martensite with increasing temperature However, above 440~ similar

trends are also observed for the chi phase formation Small differences in chemical com-

position (lower chromium and molybdenum contents, higher carbon content) could also

favor the structural stability of the martensitic regions and play a role in their relative swelling

resistance

As for the EM10 alloy, the irradiation of samples up to a maximum dose of 40 dpa did

not produce any voids Neutron irradiations up to higher doses are still necessary to establish

the limits of this void swelling resistance However, 1 MeV simulation experiments [13]

have already shown that in spite of the absence of stabilizing elements such as niobium and

vanadium, the swelling behavior of EM10 is similar to that observed in the martensitic

regions of EM12 Consequently, the fully martensitic structure that is retained even at high

irradiation temperature should imply a higher swelling resistance than for EM12 Moreover,

the chemical composition of the EM10 alloy (9 Cr-1 Mo) avoids the structural instabilities

(chromium-rich a ' ferrite in F17 and chromium-molybdenum-rich chi phase in EM12) that

have been observed to correlate with the void swelling phenomenon

Effect o f Irradiation on Mechanical Properties

The tensile and impact tests performed on the F17 steel have shown that its mechanical

properties were severely degraded by irradiation However, aging also leads to significant

embrittlement of this alloy, and the single common feature between the aged and the

irradiated material is a ' formation This phase has already been recognized as responsible

for the "475~ embrittlement" of the aged chromium-rich ferritic steels [2] The enhance-

ment of et' nucleation and growth by irradiation (shown above) partly explains the additional

embrittlement observed in irradiated F17 steel as compared to aged material Moreover,

the results obtained by G r o b n e r [2] have shown that the ~' nucleation was the most important

process with respect to decreased toughness, while substantial growth of a ' particles was

30 EFFECTS OF RADIATION ON MATERIALS

FIG 14 Comparison of the microstrucmre observed i~ aged (left) cmd irradialed (right) FI 7 alloy

also necessary to induce a severe reduction in elongation These results are in agreement

with our own observations, since a significant effect of dose has been detected in the tensile

properties of the F17 steel irradiated below 420~ and not in impact properties However,

at the lowest temperature, the irradiation also induced a high density of dislocations in the

recrystallized F17 alloy To explain the shift in embrittlement of irradiated FI7 towards low

temperatures, both the high dislocation density and its immediate locking by c~' particles

must be taken into account

GILBON ET AL ON BEHAVIOR AND MICROSTRUCTURE 31

32 EFFECTS OF RADIATION ON MATERIALS

FIG 15 Di.sloc'atio:z :~z&.:.:).s::.i~e:~:,e it7 g/Te, ir:'~te{i~ted f : 7 a:zd E "~[J:) s te,.':~

Understanding the mechanical properties of the EM12 alloy' is more difficult, since the

impact tests were performed on specimens irradiated at 40 dpa, while TEM examinations

were only made on a few tube samples irradiated up to 10/~ dpa The chi phase does not

seem responsible for the degradation of the mechanical properties of the EM12 alloy, since

maximum in USE Similar observations have been made by Gelles [5] for the irradiated

HT9 steel (12Cr-lMo VW), and the low temperature embrittlement of this alloy was at-

tributed to a fine dispersion of ~' and G phases In the EM12 steel, such phases are unlikely

GILBON ET AL ON BEHAVIOR AND MtCROSTRUCTURE 3 3

to occur since its c h r o m i u m and nickel contents are lower than for HT9, but fine dispersions

of vanadium carbide (V~C3) or Laves phase (FezMo or FezNb) could explain the embrittle- ment of this alloy [6,12.14-17] Further experiments are still necessary to identifv the nature

of the different phases formed in EM12 and to d e t e r m i n e their respective effects on me- chanical properties

significant degradation of the mechanical properties This result can again be attributed to the higher stability of this alloy, where the effect of irradiation up to 40 dpa is only to induce

a small a m o u n t of hardening at low t e m p e r a t u r e due to the increased dislocation density and the formation of small loops

Summa~

This work has covered analysis of the irradiation's effect on void swelling and mechanical properties for three ferritic steels, namely F17 EM12, and E M I 0 As with most of the ferritic steels, the 17-Cr F17 alloy is highly swelling resistant, but irradiation also enhances

a ' phase formation, which leads to severe e m b r i t t l e m e n t , increasing as the irradiation tem- perature is reduced These features make the F17 steel inappropriate for structural in-core applications in fast b r e e d e r reactors

To a lesser extent, the mechanical properties of the duplex EM12 steel are also d e g r a d e d

by irradiation, and finally, the martensitic steel EM 10 (9 Cr-1Mo) is by far the most attractive candidate for wrapper applications The initial results obtained on this alloy ( < 4 0 dpa) still need to be confirmed by further irradiations to higher doses H o w e v e r the fully martensitic structure of the E M 1 0 steel suggests a high swelling resistance, while its chemical composition should inhibit the microstructure instabilities which are responsible for the e m b r i t t l e m e n t

of the F17 and EM12 steels

Acknowledgments

The authors wish to thank MM G Allegraud, P Coffre, and R Schauff for all the mechanical tests p e r f o r m e d in the hot cells of Saclay, and MM C Rivera, O Rabouille, and P G r o s j e a n for their important assistance in the area of electron microscopy

References

[I] Allegraud, G., Boutard J L., Borer, J M., Cauvin, R Daniel, R., and Grivaud A "Em- brittlement of a 17Cr Ferritic Steei Irradiated in Ph6nix" Proceedings or the International Con- ference on Materials for Nuclear Reactor Core Application British Nuclear Energy Society London,

of Mechanical Engineers New York, 1984 pp 559-568

[6 t Gelles, D S., "'Microstructural Examination of Several Commercial Alloys Neutron Irradiated to

100 dpa," Journal o]" Nuclear Materials, Vol 148, 1987, pp 136-144

[7] Little, E A., "'Void Swelling in Irons and Ferritic Steels, I: Mechanisms of Swelling Suppression."

Journal of Nuclear Materials Vol 87 1979, pp 11-24