当前位置：文档库 › Development of Oxide Dispersion

Development of Oxide Dispersion

Development of Oxide Dispersion Strengthened Steels for High

Temperature Nuclear Structural

Applications

Hanliang Zhu,Tao Wei,Robert Harrison,Lyndon Edwards

and Kouichi Maruyama

Abstract Oxide dispersion strengthened(ODS)steels are the most promising candidate materials for high temperature nuclear applications.Mechanical alloying and subsequent thermomechanical treatments are applied to manufacture the ODS steels.Recently improved chemical composition and manufacturing processes have been developed to produce ultra?ne grain size with high number-density of nanoscale oxide particles and high dislocation density in the https://www.wendangku.net/doc/596986087.html,ually,?ne grains degrade creep resistance at elevated temperatures.However,the?ne-grained ODS steels exhibit not only good radiation resistance,but also superior creep properties.The present paper reviews the chemical compositions,manufacturing processing,microstructural features,thermal creep properties and radiation resis-tance of recently developed ODS steels.Special attention is paid to the effects of the ?ne-scale microstructural features on thermal creep and radiation resistance. Keywords Oxide dispersion strengthened steelsáCreep resistanceáRadiation resistance

1Introduction

The structural materials for Generation IV(GenIV)and future fusion power reactors will need to endure higher temperatures and more severe radiation exposure than in

H.Zhu(&)áT.WeiáR.HarrisonáL.Edwards

Institute of Materials Engineering,Australian Nuclear Science and Technology

Organisation,Locked Bag2001,Kirrawee DC,Sydney,NSW2232,Australia

K.Maruyama

Graduate School of Environmental Studies,Tohoku University,02Aobayama,

Sendai980-8579,Japan

J.Mathew et al.(eds.),Engineering Asset Management and

1147 Infrastructure Sustainability,DOI:10.1007/978-0-85729-493-7_89,

óSpringer-Verlag London Limited2011

1148H.Zhu et al. commercial reactors operating today.For example,the core structural materials for GenIV reactors will operate at500–1,000°C and experience damage up to*30–100 displacements per atom(dpa),while today’s reactors experience temperatures up to 400°C and damage up to20dpa[1].The more extreme operating conditions for future nuclear power reactors are beyond those experienced in current nuclear power plants,and require the development of new high-performance materials.

Reduced-activation ferritic/martensitic(F–M)steels are currently the leading structural materials for core components of future nuclear reactors[2].The reduced activation steels are those without the addition of conventional alloying elements such as Nb,Mo,Co,Ni,Cu,N etc.[3,4].These steels show rapid radioactive decay after neutron irradiation,allowing shallow burial of the com-ponents after service[4,5].Also,these steels exhibit signi?cantly better radiation resistance and superior mechanical properties compared with commercial steels [3,5].However,the irradiation embrittlement at low temperatures(\400°C)and inferior long-term creep rupture strength at high temperatures limit their appli-cations in future nuclear systems[5].

An important way to improve the thermal creep resistance and radiation resistance is through the addition of?ne oxide particles to form so-called oxide dispersion strengthened(ODS)steels[4].By optimizing chemical compositions and manufacturing processes,?ne grain size and nanoscale oxide particles can be produced in the microstructure[4,6–8].It is well known that a?ne grain size normally degrades the thermal creep strength at elevated temperatures[9].How-ever,the?ne-grained nano-structured ODS steels exhibit not only good radiation resistance,but also excellent creep resistance[6,10].So far,it is not fully understood why the?ne-grained ODS steels can still have good creep properties and can be used at high operating temperatures.The present paper reviews the chemical compositions,manufacturing processing,microstructural features,ther-mal creep properties and radiation resistance of recently developed ODS steels. The effects of?ne grain size and nanoscale oxide particles on thermal creep and radiation resistance are specially discussed.

2Compositions,Processing and Microstructure

of ODS Steels

In steels,Chromium is a ferrite stabilising element.According to the Fe–Cr constitution diagram,there is a c loop[11].At above approximately13%Cr,the binary Fe–Cr alloys are ferritic over the whole temperature range.At about9%Cr, there is an extensive austenitic region from820to1200°C and the two-phase region of austenite and ferrite has a very narrow temperature range.It is possible to produce a fully martensitic microstructure with minimal amounts of ferrite phase that is generally regarded as detrimental to high temperature strength properties [11].Hence,the optimum concentration of Cr from the point of view of creep strength and toughness for heat-resistance steels was thought to be9%[11,12].

Development of Oxide Dispersion Strengthened Steels1149 However,for ODS alloys as candidate nuclear structural materials,radiation resistance is another important property to be considered.BCC ferritic alloys exhibit better radiation resistance than FCC austenitic alloys[4,10].Increasing the Cr content can increase the resistance to irradiation-induced swelling and improve the corrosion–oxidation resistance[10,13].Hence,both9%Cr and[12%Cr ODS steels have been developed in recent years.Table1lists the chemical composi-tions of typical ODS steels developed in different countries.Among the chemical elements added to these steels,C is a strong austenite stabilising element,and has a very small solubility in ferrite,resulting in the formation of carbides for precipi-tation strengthening[14].Tungsten and Mo can provide solid-solution strength-ening[15].Yttria(Y2O3)is added to introduce the oxide particles for dispersion strengthening[10].The addition of small amount of Ti can further reduce the size of the oxide particles[16].Manganese and Ni can be used for phase control[15]. Tantalum is sometimes added as a replacement for Nb for forming stable carbides [4,14].Aluminium can improve the corrosion resistance,but can cause coarsening and inhomogeneous dispersion of the oxide particles,suggesting that it should be strictly controlled or eliminated from the ODS steels[17,18].

The ODS alloys can be produced by conventional casting processes or mechanical alloying(MA).The former method produce alloys with non-uniform distribution of dispersoids,while the latter is a powder processing route by which uniform distribution of oxide particles can be achieved throughout the matrix[19]. Figure1illustrates a typical fabrication route for an ODS steel[10,20,21].The rapid solidi?ed alloy and ultra?ne oxide powder are?rst mixed by ball milling. The as-milled powders are then consolidated by hot isostatic pressing or hot extrusion.Thermomechanical treatments(TMT),involving annealing heat treat-ments,hot and/or cold rolling,and various heat treatment processes,are conducted to the extruded materials to fabricate products such as test samples,tube,sheet, and plate etc.,and to obtain designated mechanical properties.

Ball milling deforms and mixes the steel and oxide powders,resulting in an enormous supersaturation of dissolved Y and O,the formation of?ne grain size and high dislocation densities in the as-milled powders[10].Hot extrusion and/or hot rolling produce a?nely elongated and textured grain structure parallel to the extrusion or rolling direction[10,22].For example,the microstructure of a12Cr-ODS sheet had10.0l m grain width and1.0mm grain length,resulting in a length-to-width ratio of100,and a strong texture of{001}\110[[22].Such textures and elongated grain structures cause anisotropic mechanical properties and low fracture toughness[10].The post-consolidation TMT treatments can change the grain morphology by deformation,phase transformation and recrys-tallization.For12Cr ODS alloys,adequate annealing heat treatments following cold rolling produces a fully recrystallised ferritic microstructure,and the recrystallisation process results in coarse grains with a lower aspect ratio[15,21, 23].For9Cr ODS alloys,a to c phase transformation was used to produce an equiaxed tempered martensite with d-ferrtic phase[15,23].By using optimal milling time and decreasing extrusion temperature to850°C in a14YWT alloy, ultra?ne grain size of the order of100nm could be produced[6,8].The average

T a b l e 1C h e m i c a l C o m p o s i t i o n o f V a r i o u s O D S S t e e l s (w t %)

A l l

o y C r

C T i W M n M o A l N i V T a F e O x i d

e O t h e r R e f

M A 956200.050.5–––4.5––B a l .0.5[16]M A 95714–0.9––0.3–––B a l .0.25[16]12Y W T 12.

60.050.352.40.05––0.3–B a l .0.25[16]14Y

W T 14–0.43–––––B a l .0.3[46]P M 200020\0.04

0.5–––5.5

––B a l .0.5[47]D T 2203Y O

3–2.2––1.5–––B a l .0.5[48]E u r

o f e

r 9

78.90.11–1.10.47–––0.20.

B a l .0.3[49]M B 18A

9.040.130.21.930.04––0.02–B a l .0.36[50]M m 159.020.140.201.99\0.01––0.01–B a l .0.36[50]T 38.750.130

.93

––

0.01–

[50]E P 4

13.

.71

.88

1B :0.004N b :0.25

P :0.018[51]1150

H.Zhu et al.

Development of Oxide Dispersion Strengthened Steels1151

grain lengths in the transverse and longitudinal extrusion direction were110and 150nm,respectively[6].Also,the grain structure was highly stable in that the grains did not grow signi?cantly after annealing for several hours at1,000°C[6].

The oxide particles precipitate in the ODS steels during consolidation at ele-vated temperatures[10,16,24].At least two types of oxide phases were found in a commercial MA957alloy:non-stoichiometric Y-,Ti-,O-enriched clusters from 2to15nm(Y/Ti\1)and stoichiometric Y2Ti2O7from15to35nm[25].Ti,Y and excess oxygen concentration are critical for the?nal morphology of the oxide particles in the ODS steels.Titanium decreased the oxide particle diameter from more than10to3nm in a12Cr-ODS alloy[26].The additions of Y2O3to an Fe—14wt%Cr—3%W—0.4%Ti(14WT)alloy decreased the size of the oxide particles from4to2nm,increased the number density of oxide particles by a factor of approximately10,and greatly increased the thermal stability of the oxide particles[7].A certain amount of excess oxygen concentration,which is de?ned as the value obtained by subtracting the oxygen concentration coupled with yttrium (in the form of Y2O3)from the total oxygen concentration,was found to be important in in?uencing the size and distribution of the oxide particles in the ODS steel[27].Additionally,the process parameters such as milling time and heat treatment temperature are important factors that in?uence the?nal morphology of the oxide particles[8].It was found that when the extrusion temperature was decreased from 1,175to850°C for the14YWT alloy,the oxide particles changed from20nm Y2O3to *4nm Y-,Ti-,O-enriched clusters[7,8].Furthermore,the nanoscale oxide particles, which were referred to as nanoclusters by the authors,in the recently developed 14YWT ODS alloy were found to be extremely stable at temperatures of at least 1,000°C[7].There was also a much higher number-density of nanoscale particles at the grain boundaries,*791024m-3,than in the grain interiors*391023m-3,while the diameter of the oxide particles on the grain boundaries,2.8nm,was only slightly larger than that in the grain interiors,2.4nm[6].

3Creep Resistance of ODS Steels

The creep deformation behaviour and deformation mechanisms of ODS steels have been investigated.Figure2illustrates a Larson–Miller plot comparing the creep strength of typical ODS steels and those of similar steels without the oxide particles.It can be observed that the ODS steels exhibit better creep resistance than the non-ODS materials.The creep resistance of metallic materials is strongly dependent on their microstructure[9].

The anisotropic grain microstructure in ODS steels has a signi?cant effect on creep properties [21,23,28,29].The unrecrystallised specimens with ?nely elongated grains in the extrusion direction exhibit strength anisotropy [14,21,23].The creep rupture strength in the hoop direction of an unrecrystallised 13Cr-ODS steel was three times lower than in the uniaxial direction [21].A fully recrys-tallisation treatment improved the creep rupture strength of the hoop direction to close to that of the uniaxial direction in the 13Cr-ODS steel [21],and resulted in no difference in the two directions of a 13Cr-ODS (1DK-ODS)steel [23].The suppression of grain boundary sliding was suggested to be responsible for improved creep rupture strength in the hoop direction of the recrystallised spec-imens [22,28,29].

As shown in Fig.2,the ultra?ne-grained 14YWT also exhibits excellent creep resistance.Fine grains usually degrade the creep strength due to accelerated dif-fusion by grain boundaries at elevated temperatures above a homologous tem-perature T/T melting of about 0.4.The effect of grain size on creep deformation behaviour was investigated for the 14YWT [6].It was found that even at 800°C,which corresponds to a homologous temperature of 0.6,the ?ne-grained material with a grain size of about 200nm showed higher creep strength than the coarse-grained one with a grain size of approximately 250l m.When the deformation temperature as increased to 950°C and above,the creep strength of the ?ne-grained material dropped lower than that for the coarse-grained.Also,it was found that at the deformation temperature of 800°C,the stress exponent decreased from a value of 35at high stresses to a much smaller value on the order of 1at low stresses.With an increase in the deformation temperature to 950°C,the ?ne-grained material showed almost the same strain rate as in the coarse-grained one.However,the stress exponent for the coarse-grained material,60,was much higher than that for the ?ne-grained,13.This is related to the morphology and distribution of the oxide particles and their effect on deformation mechanisms.

The deformation mechanism map for ODS steels has not yet been established,however,the deformation mechanisms of ODS steels can be discussed on the

basis Fig.2Larson–Miller plot of

various ODS steels 1152

H.Zhu et al.

of the deformation map of stainless steels.Figure 3presents a deformation mechanism map of a 316stainless steel [30].Here,several deformation regimes are depicted as a function of temperature.Dislocation glide (Region A)and power-law creep (Region B)are the dominant deformation mechanism at high stresses.With decreasing stress,the dominant deformation mechanism at high temperatures can change to diffusion creep including boundary diffusion,that is,Coble creep (Region C)and bulk diffusion,that is,Nabarro–Herring creep (Region D).The core structural materials for current reactors and Gen-IV reactors operate at up to *400°C and up to *1,000°C,respectively [1].Hence,for steels applied in current nuclear reactor systems,dislocation creep is the dominant deformation mode over a wide temperature range.Dislocation creep involves the movement of disloca-tions through the crystal lattice of the material.Any microstructural features that can retard the motion of dislocations can improve the creep resistance.The oxide particles in the ODS materials provide and persistent strong barriers to dislocation motion [13],resulting in better creep resistance in the ODS steels than non-ODS materials as shown in Fig.2.

However,for ?ne-grained materials,grain boundary sliding (GBS)can also be an important deformation mechanism of plastic ?ow at high homologous tem-peratures [31].GBS during creep of polycrystals is considered to be an inde-pendent deformation mechanism which follows speci?c relations involving stress,temperature,strain rate and grain size [31,32].The deformation mechanism map of Fig.3does not include grain boundary sliding (GBS)as a deformation mechanism.According to Refs.[31]and [32],the domains of bulk diffusion-and grain boundary diffusion-controlled GBS can be added among power-law (Region B),Cobel creep (Region C)and Nabarro–Herring creep domains for ?ne-grained materials.With decreasing grain size,the domains of GBS can extend to low temperatures and low stresses due to increasing the number of grain boundaries [33].Deformation by diffusional ?ow mechanisms will only become important when very low stresses,e.g.r B 10-5E ,are achieved for a stainless steel [31

Fig.3Creep deformation

map for a 316stainless steel

[30].A—dislocation glide,

B—power-law creep,C—

boundary diffusion,D—bulk

diffusion,and E—elastic

regime Development of Oxide Dispersion Strengthened Steels

1153

The strength of the grain boundary and grain matrix change at different rates with temperature,and are in?uenced by second-phase particles.Figure 4illustrates schematically the strength of the grain boundary and grain interior in metallic materials as a function of temperature and second-phase particles in the case that the particles strengthen the materials.At low temperatures,the grain boundaries are stronger than the matrix,and can act as a barrier to deformation.With increasing deformation temperature,diffusion is accelerated at the grain bound-aries,and the boundaries are softer than the matrix.A crossover temperature Tc can be de?ned where the role of the grain boundary is changed [33,34].Hence,grain size has the opposite effect on the ?ow stress at high and low temperatures.If the deformation temperature is lower than the crossover temperature,grain boundaries can strengthen the materials,and the ?ner-grained materials exhibit superior creep resistance than the coarser-grained ones due to higher numbers of grain boundaries.However,if the deformation temperature is higher than the crossover temperature,the role of grain boundaries can change to soften materials,and the coarser-grained materials show better creep resistance than the ?ner-grained ones.In addition,the strength of materials as well as the crossover tem-perature is also in?uenced by the deformation rate.The discussion in the present paper is limited to the creep rates \10-3s -1.

At high deformation temperatures above the crossover temperature,grain boundary sliding can easily become the dominant deformation mechanism for ?ne-grained materials,while only elastic deformation occurs for coarse-grained materials at low stresses less than a certain level.Hence,?ne grains do decrease creep resistance of materials at certain temperature above the crossover temper-ature and certain stress range;this has been con?rmed in various materials [9,35].Also,for the 14YWT,the reported result that the creep strength of a ?ne-grained sample is lower than that of coarse-grained one at over 950°C and the

stress Fig.4Strength of the grain

boundary and grain matrix as

a function of temperature and

second-phase particles 1154

H.Zhu et al.

Development of Oxide Dispersion Strengthened Steels1155 exponent of the?ne-grained one is much lower than that of the coarse-grained one at high applied stress and950°C[6]indicate that?ne grains do decrease the creep resistance of the ODS steels.

The presence of a high number density of second-phase particles can increase the strength of both the grain boundary and interior due to pinning of dislocation motion and inhibiting diffusion,and can increase the crossover temperature from T c1to T c2as shown in Fig.4.Hence,the second-phase particles can increase the service temperature for the metallic materials.It was reported that the upper temperature limit in thermal creep strength could be enhanced at least100°C by introducing the nanoscale oxide particles in ODS steels[9,36].However,if the second-phase particles are unstable,such as Fe2W laves phase in a9Cr-W steel [37],they could become coarser during creep at elevated temperatures,reducing the precipitation strengthening and promoting the acceleration of creep rate in the tertiary or acceleration creep region after reaching a minimum creep rate.As described earlier,the nanoscale oxide particles in14YWT are extremely stable, maintaining their strengthening effect during long-term creep.Also,the oxide particles at the grain boundaries had a much higher number density and only slightly larger diameter than those in the grain interiors[6].These nanoscale oxide particles at the grain boundaries could effectively pin diffusion and sliding of grain boundaries at high temperatures,maintaining the strengthening effect of the grain boundary region.Therefore,the?ne-grained material could still exhibit better creep resistance than the coarse-grained one even at the homologous temperature of0.6.However,with a further increase in temperature above a certain level (950°C for14YWT[6]),the strengthening effect of the oxide particles becomes so weak that the grain boundary region becomes softer than the grain matrix again, and the?ne-grained materials exhibits lower creep strength than that of the coarse-grained[6].Hence,the presence of nanoscale oxide particles at the grain boundary region decreases the negative effect of?ne grain size on creep to some extent. Because most creep tests of ODS steels in the published references were carried out near its target operating temperature,more experimental work on creep at different deformation temperatures is needed in order to more fully understand the deformation behaviour and deformation mechanisms of?ne-grained ODS materials.

4Radiation Resistance of ODS Steels

When a material is irradiated,neutrons can displace metal atoms from their normal lattice positions,resulting in the formation of vacancies and interstitials.Neutrons can also be absorbed by the atoms of the irradiated material,producing new species by transmutation and gas atoms of hydrogen and/or helium within the matrix[38].The primary defects can further migrate by diffusion to form defect clusters such as dislocation loops,voids,bubbles and precipitates.These micro-structural evolutions result in changes in mechanical properties and dimensions.

1156H.Zhu et al. Hence,any microstructural features that can retard the motion of the primary defects can suppress the accumulation of defect clusters,and enhance radiation resistance.The oxide particles in the ODS steels are one of such microstructural features.

ODS steels have been irradiated under different environments including ion beams,electrons,and neutrons[10,39,40].A large number of studies indicate that the oxide particles are stable under irradiation.However,recently,a focused study on the stability of the oxide particles during irradiation found that the average size of nanoscale oxide particles in a9Cr-ODS ferritic steel decreased with increase of density after heavy ion irradiation at500–700°C to a dose of150dpa[39].The mean oxide diameter and the number density changed from11.7nm and 3.491021m-3before irradiation to4.9nm and1.691022m-3,respectively, after irradiation.The increased number density of the oxide particles should give a dynamic strengthening to the steel matrix during creep deformation[39], improving the creep resistance.In addition,the dislocations and grain structures were found to undergo only minor changes in MA957ODS steels after neutron irradiation at*750°C and*100dpa[10].

The high number-density and nanoscale of the oxide particles in these ODS steels are effective traps for helium atoms and/or vacancies by offering a high number of trapping sites,accelerating nucleation of bubbles and preventing the voids from growing,hence resulting in superior resistance to swelling[41,42]. Dislocation loops and bubble–loop complexes were observed in a PM2000ODS steel implanted by He to approximately0.75dpa at300–500°C[42],and cavities was found in a16Cr-ODS steel with He to approximately60dpa at500°C[41]. However,the average size and number density of cavities formed in the16Cr-ODS steel were half the size and twice the density of those in a9Cr-2W steel without the presence of the oxide particles,indicating that the introduction of the oxide particles did improve the resistance to swelling.It was also found that the matrix of an MA957ODS steel had smaller and more numerous bubbles than that of a non-ODS F82H after irradiation at500°C to approximately9dpa and380appm He, while the grain boundaries of the ODS steel had a lower concentration of bubbles than that of F82H[10].All these results indicated that the oxide particles can re?ne the defect clusters and could result in a homogeneous distribution of defect clusters due to irradiation.

The ODS steels can also show excellent resistance to irradiation hardening due to the presence of the oxide particles.For example,both irradiation hard-ening and the accompanied ductility losses were lower in an ODS steel,MA957, than conventional TMS alloy without the oxide particles[10,43].It was summarised in Ref.[10]that irradiation hardening increased with decreasing irradiation temperatures below approximately550°C but was minimal at higher temperatures,and hardening was due primarily to dislocation loops and the formation of Cr-rich a precipitates below and above approximately400°C, respectively[17].For the recently developed14YWT alloy,a slight increase in yield strength(10–125MPa)at room testing temperature was found after irra-diation at temperatures of300,580and670°C[36],while the increase of yield

Development of Oxide Dispersion Strengthened Steels1157 strength was about250MPa at irradiation temperature of300°C for the14WT without the addition of the oxide particles,again indicating that the addition of oxide particles decreased the irradiation hardening at low temperatures for 14YWT.There was no evident difference in ductility change before and after irradiation at300°C for both14WT and14YWT alloys.In the same research,it was found that an ODS-EUROFER steels with relatively larger grains and lower density of oxide particles than the14YWT showed an increase in yield strength of277MPa and decrease in total elongation of about4.5%after irradiation at 300°C[36].The?ne grain size and small size with high number density oxide particles in14YWT should be responsible for the unremarkable irradiation hardening behaviour in this material.

The effect of oxide particles on irradiation creep is not as signi?cant as on thermal creep.Irradiation creep in MA957was approximately athermal between 400and600°C[10],while irradiation creep rates of a19Cr-ODS steel showed linear stress dependence up to250MPa at temperatures from300to500°C[44]. For MA957,the irradiation creep compliance,Ic,ranged from approximate 0.8910-6to0.25910-6/MPa-dpa,which is slightly lower than in tempered martensitic steels(TMS)and is signi?cantly smaller than in austenitic stainless steels[10].Also,the irradiation creep compliance increased with increasing testing temperature.For example,the irradiation creep compliance was found to increase from4.0910-6to11910-6/MPa-dpa for a19Cr ODS steel,and5.7910-6to 18910-6/MPa-dpa for a PM2000when the irradiation temperature increased from300to500°C[44].Furthermore,the size and distribution of the oxide par-ticles did not show a signi?cant in?uence on the irradiation creep behaviour, inferring that irradiation creep is a pure matrix phenomenon[44,45].More experiment work and theoretical analysis are needed for fully understanding the irradiation behaviour in the nano-structured ODS steels at different irradiation conditions.

5Summary

Recent developments of chemical compositions and manufacturing processes for ODS steels have resulted in microstructures with ultra?ne grain size and ultrahigh density of stable nanoscale oxide particles.The?ne microstructural features can effectively act as sinks to absorb primary defects during irradiation,increasing radiation resistance.Furthermore,high number-density,stable and nanoscale particles can retard the motion of dislocations during creep deformation, increasing thermal creep resistance.Fine grain size can increase creep properties at low operating temperatures while it can decrease creep resistance at high tem-peratures due to accelerated diffusion and GBS.However,the nanoscale oxide particles at the grain boundary region can effectively pin the diffusion and GBS, decreasing the harmful effect of?ne grain size on creep at high temperatures to some extent.

1158H.Zhu et al. References

1.Development of radiation resistant reactor core structural materials.https://www.wendangku.net/doc/596986087.html,/

About/Policy/GC/GC51/GC51InfDocuments/English/gc51inf-3-att7_en.pdf.Accessed29 Sept2010

2.Kurtz RJ,Alamo A,Lucon E,Huang Q,Jitsukawa S,Kimura A,Klueh RL,Odette GR,Petersen

C,Sokolov MA,Sp?tig P,Rensman JW(2009)Recent progress toward development of reduced activation ferritic/martensitic steels for fusion structural applications.J Nucl Mater386–388: 411–417

3.Zinkle SJ(2005)Advanced materials for fusion technology.Fusion Eng Des74:31–40

4.Allen TR,Busby JT,Klueh RL,Maloy SA,Toloczko MB(2008)Cladding and duct materials

for advanced nuclear recycle reactors.JOM60(1):15–23

5.Murty KL,Charit I(2008)Structural materials for Gen-IV nuclear reactors:challenges and

opportunities.J Nucl Mater383:189–195

6.Schneibel JH,Liu CT,Miller MK,Mills MJ,Sarosi P,Heilmaier M,Sturm D(2009)

Ultra?ne-grained nanocluster-strengthened alloys with unusually high creep strength.Scripta Mater61:793–796

https://www.wendangku.net/doc/596986087.html,ler MK,Russell KF,Hoelzer DT(2006)Characterization of precipitates in MA/ODS

ferritic alloys.J Nucl Mater351:261–268

8.West MK(2006)Processing and characterization of oxide dispersion strengthened14YWT

ferritic alloys.Ph.D thesis,The University of Tennessee,Knoxville

9.Kassner ME,Perez-Prado MT(2004)Fundamentals of creep in metals and alloys.Elsevier

Ltd,Oxford

10.Odette GR,Alinger MJ,Wirth BD(2008)Recent developments in irradiation-resistant steels.

Ann Rev Mater Res38:471–503

11.Ennis PJ,Czyrska-Filemonowicz A(2002)Recent advances in creep resistant steels for

power plant applications.Power Plant:Opera Maintence Mater Issues1(1):1–28

12.Abe F(2008)Precipitate design for creep strengthening of9%Cr tempered martensitic steel

for ultra-supercritical power plants.Sci Technol Adv Mater9:013002

13.Hayashi T,Sarosi PM,Schneibel JH,Miller MJ(2008)Creep response and deformation

processes in nanocluster-strengthened ferritic steels.Acta Mater56:1407–1416

14.Klueh RL(2005)Elevated temperature ferritic and martensitic steels and their application to

future nuclear reactors.Int Mater Rev50:287–310

15.Inoure M,Kaito T,Ohtsuka S(2007)Research and development of oxide dispersion

strengthened ferritic steels for sodium cooled fast breeder reactor fuels.NATO advanced study institute on materials for generation-IV nuclear reactors,Institut d’Etudes Scienti?ques de Cargese—Cargese,CORSICA,France

16.Ei-Genk MS,Tournier JM(2005)A review of refractory metal alloys and mechanically

alloyed-oxide dispersion strengthened steels for space nuclear power systems.J Nucl Mater 340:93–112

17.Oh S,Lee JS,Jang C,Kimura A(2009)Irradiation hardening and embrittlement in high-Cr

oxide dispersion strengthened steels.J Nucl Mater386–388:503–506

18.Takaya S,Furukawa T,Aoto K,Muller G,Weisenburger A,Heinzel A,Inoue M,Okuda T,

Abe F,Ohnuki S,Fujisawa T,Kimura A(2009)Corrosion behavior of Al-alloying high Cr-ODS steels in lead-bismuth eutectic.J Nucl Mater386:507–510

19.Susila P,Sturm D,Heilmaier M,Murty BS,Sarma VS(2010)Microstructural studies on

nanocrystalline oxide dispersion strengthened austenitic(Fe-18Cr-8Ni-2W–0.25Y(2)O(3)) alloy synthesized by high energy ball milling and vacuum hot pressing.J Mater Sci45: 4858–4865

https://www.wendangku.net/doc/596986087.html,ai S,Mizuta S,Fujiwara M,Okuda T,Kobayashi T(2002)Development of9Cr-ODS

martensitic steel claddings for fuel pins by means of ferrite to austenite phase transformation.

J Nucl Sci Technol39:778–788

Development of Oxide Dispersion Strengthened Steels1159 https://www.wendangku.net/doc/596986087.html,ai S,Nishida T,Okada H,Okuda T,Fujiwara M,Asabe K(1997)Development of oxide

dispersion strengthened ferritic steels for FBR core application(I).Improvement of mechanical properties by recrystallization processing.J Nucl Sci Technol34:256–263

22.Okada H,Ukai S,Inoue M(1996)Effects of grain morphology and texture on high

temperature deformation in oxide dispersion strengthened ferritic steel.J Nucl Sci Technol 33:936–943

https://www.wendangku.net/doc/596986087.html,ai S,Nishida T,Okuda T,Yoshitake T(1998)R&D of oxide dispersion strengthened

ferritic martensitic steels for FBR.J Nucl Mater258–263:1745–1749

24.Alinger MJ,Odette GR,Hoelzer DT(2009)On the role of alloy composition and processing

parameters in nanocluster formation and dispersion strengthening in nanostuctured ferritic alloys.Acta Mater57:392–406

25.Sakasegawa H,Chaffron L,Legendre F,Boulanger L,Cozzika T,Brocq M,Carlan Y(2009)

Correlation between chemical composition and size of very small oxide particles in the MA957ODS ferritic alloy.J Nucl Mater384:115–118

https://www.wendangku.net/doc/596986087.html,ai S,Fujiwara M(2002)Perspective of ODS alloys application in nuclear environments.

J Nucl Mater307–311:749–757

https://www.wendangku.net/doc/596986087.html,ai S,Ohtsuka S(2007)Nano-mesoscopic structure control in9Cr–ODS ferritic steels.

Energy Mater Mater Sci Eng Energy Syst2(1):26–35

https://www.wendangku.net/doc/596986087.html,ai S,Mizuta S,Yoshitake T,Okuda T,Fujiwara M,Hagi S,Kobayashi T(2000)Tube

manufacturing and characterization of oxide dispersion strengthened ferritic steels.J Nucl Mater283–287:702–706

https://www.wendangku.net/doc/596986087.html,ai S,Okuda T,Fujiwara M,Kobayashi T,Mizuta S,Nakashima H(2002)Characterization

of High Temperature Creep Properties in Recrystallized12Cr–ODS Ferritic Steel Claddings.

J Nucl Sci Technol39(8):872–879

30.Frajtag P,Charit I,Murty KL(2007)Structural materials for the next generation nuclear

reactors—an overview.2007International Congress on Advances in Nuclear Power Plants, Nice Acropolis,France,https://www.wendangku.net/doc/596986087.html,/icapp07/MONDAY/KN-B/7315Gen%20IV% 20Presentation-?nal.pdf.Accessed20Sep2010

31.Ruano O,Wadsworth J,Sherby OD(1985)Deformation mechanisms in an austenitic

stainless steel(25Cr-20Ni)at elevated temperature.J Mater Sci20:3735–3744

32.Luthy H,White RA,Sherby OD(1979)Grain boundary sliding and deformation mechanism

maps.Mater Sci Eng A Struct Mater Prop Microstruc Process39:211–216

33.Pan J,Tong J,Tian M(2002)Fundamental of Materials Science.Qinghua University

Publisher,Beijing

34.Zhu H,Dahle AK,Ghosh AK(2009)Effect of Sc and Zn additions on microstructure and hot

formability of Al-Mg sheet alloys.Metallurg Mater Trans A40(3):598–608

35.Meyers MA,Mishra A,Benson DJ(2006)Mechanical properties of nanocrystalline

materials.Prog Mater Sci51:427–556

36.McClintock DA,Sokolov MA,Hoelzer DT,Nanstad RK(2009)Mechanical properties

of irradiated ODS-EUROFER and nanocluster strengthened14YWT.J Nucl Mater392: 353–359

37.Abe F(2005)Effect of?ne precipitation and subsequent coarsening of Fe2W laves phase on

the creep deformation behavior of tempered martensitic9Cr-W steels.Metallur Mater Trans A36:321–332

38.Little EA(2006)Development of radiation resistant materials for advanced nuclear power

plant.Mater Sci Technol22(5):491–518

39.Allen TR,Gan J,Cole JI,Miller MK,Busby JT,Shutthanandan S,Thevuthasan S(2008)

Radiation response of a9chromium oxide dispersion strengthened steel to heavy ion irradiation.J Nucl Mater375:26–37

40.Nanstad RK,McClintock DA,Hoelzer DT,Tan L,Allen TR(2009)High temperature

irradiation effects in selected Generation IV structural alloys.J Nucl Mater392:331–340 41.Yutani K,Kishimoto H,Kasada R,Kimura A(2007)Evaluation of helium effects on swelling

behavior of oxide dispersion strengthened ferritic steels under ion irradiation.J Nucl Mater 367–370:423–427

1160H.Zhu et al.

42.Chen J,Jung P,Hoffelner W,Ullmaier H(2008)Dislocation loops and bubbles in oxide

dispersion strengthened ferritic steel after helium implantation under stress.Acta Mater 56:250–258

43.Alamo A,Bertin JL,Shamardin VK,Wident P(2007)Mechanical properties of9Cr

martensitic steels and ODS-FeCr alloys after neutron irradiation at325degrees C up to42 dpa.J Nucl Mater367–370:54–59

44.Chen J,Pouchon MA,Kimura A,Jung P,Hoffelner W(2009)Irradiation creep and

microstructural changes in an advanced ODS ferritic steel during helium implantation under stress.J Nucl Mater386–388:143–146

45.Chen J,Hoffelner W(2009)Irradiation creep of oxide dispersion strengthened(ODS)steels

for advanced nuclear applications.J Nucl Mater392:360–363

46.McClintock DA,Hoelzer DT,Sokolov MA,Nanstad RK(2009)Mechanical properties of

neutron irradiated nanostructured ferritic alloy14YWT.J Nucl Mater386–388:307–311 47.Czyrska-Filemonowicz A,Dubiel B(1997)Mechanically alloyed,ferritic oxide dispersion

strengthened alloys:Structure and properties.J Mater Process Technol64:53–64

48.Evans RW,Preston JA,Wilshire B,Little EA(1993)Creep and creep fracture of an oxide-

dispersion-strengthened13-percent chromium ferritic steel.Mater Sci Eng A Struct Mater Prop Microstruct Process167:65–72

49.Lindau R,Moslang A,Schirra M,Schlossmacher P,Klimenkov M(2002)Mechanical and

microstructural properties of a hipped RAFM ODS-steel.J Nucl Mater307–311:769–772 https://www.wendangku.net/doc/596986087.html,ai S,Ohtsuka S,Kaito T,Sakasegawa H,Chikata N,Hayashi S,Ohnuki S(2009)High-

temperature strength characterization of advanced9Cr-ODS ferritic steels.Mater Sci Eng A Struct Mater Prop Microstruct Process510–511:115–120

51.Ageev VS,Vil’danova NF,Kozlov KA,Kochetkova TN,Nikitina AA,Sagaradze VV,

Safronov BV,Tsvelev VV,Chukanov AP(2008)Structure and thermal creep of the oxide-dispersion-strengthened EP-450reactor steel.Phys Metals Metallogr106(3):318–325