Atomic origins of solid helium bubbles in tungsten

Solid helium bubbles were directly observed in the helium ion implanted tungsten（W）, by different transmission electron microscopy（TEM） techniques at room temperature. The diameters of these solid helium bubbles range from1 nm to 8 nm in diameter with the mean bubble size about 3 nm. The selected area electron diffraction（SAED） and fast Fourier transform（FFT） images revealed that solid helium bubbles possess body-centered cubic（bcc） structure with a lattice constant of 0.447 nm. High-angle annular dark-field scanning transmission electron microscopy（HAADF-STEM）images further confirmed the existence of helium bubble in tungsten. The present findings provide an atomic level view of the microstructure evolution of helium in the materials, and revealed the existence of solid helium bubbles in materials.

Effect of helium implantation on SiC and graphite

Effects of helium implantation on silicon carbide（SiC） and graphite were studied to reveal the possibility of SiC replacing graphite as plasma facing materials. Pressureless sintered SiC and graphite SMF-800 were implanted with He＋ions of 20 ke V and 100 ke V at different temperatures and different fluences. The He^＋ irradiation induced microstructure changes were studied by field-emission scanning electron microscopy（FESEM）, atomic force microscopy（AFM）, and transmission electron microscopy（TEM）.

Observation of the Structure of Tungsten Films Prepared by MOCVD

The tungsten films with ultra microstructure on CuCrZr alloy and China Low Acti- vation Martensitic （CLAM） steel have been prepared by metal organic chemical vapor deposition （MOCVD）. The films were produced by pyrolysing the tungsten hexacarbonyl at air or argon atmosphere. When formed at or below 400 ℃, they were poorly crystalized and the films showed low quality in thickness, density, bonding performance etc. While above this temperature, the properties of tungsten films have been improved, all the films consist of tungsten in the β-W. And β-W can change into α-W after heat treatment. As in other variations of pyrolysis, oxy- gen and carbon were observed. When filled with argon, the oxygen and carbon content would reduce apparently. Tungsten films prepared by MOCVD have stable chemical composition and microstructure. Besides, the properties of films on CuCrZr alloy are better than that on CLAM steel.

Theoretical models for irradiation hardening and embrittlement in nuclear structural materials:a review and perspective

The study of irradiation hardening and embrittlement is critically important for the development of next-generation structural materials tolerant to neutron irradiation,and could dramatically affect the approach to the design of components for advanced nuclear reactors.In addition,a growing interest is observed in the field of research and development of irradiation-resistant materials.This review aims to provide an overview of the theoretical development related to irradiation hardening and embrittlement at moderate irradiation conditions achieved in recent years,which can help extend our fundamental knowledge on nuclear structural materials.After a general introduction to the irradiation effects on metallic materials,recent research progress covering theoretical modelling is summarized for different types of structural materials.The fundamental mechanisms are elucidated within a wide range of temporal and spatial scales.This review closes with the current understanding of irradiation hardening and embrittlement,and puts some perspectives deserving further study.

Preparation of W-TiC alloys from core-shell structure powders synthesized by an improved wet chemical method

In order to improve the homogeneous distribution of the TiC particles and facilitate the TiC particles to distribute in the tungsten grain interiors,two kinds of TiCdoped tungsten precursors with a core-shell structure were prepared by an improved wet chemical method at different reaction temperature conditions.Consequently,fine platelike precursor(200-400 nm)and flower-like precursor(approximately 1.25μm)are obtained.After reduction and sintering,the microstructures of the samples were characterized by scanning electron microscopy and transmission electron microscopy.In the sample sintered from the platelike precursor,TiC particles with sizes in the range of40-300 nm and an average size of approximately 80 nm were uniformly distributed in the tungsten matrix with a high share in the grain interiors.However,in the sample sintered from the flower-like precursor,the TiC particles with sizes in the range of 50-700 nm are significantly aggregated and non-uniformly distributed in the tungsten matrix.As a result,the sample sintered from the plate-like precursor achieves higher mechanical properties and a much narrower range of bending strength values than that sintered from the flower-like precursor.The average bending strength of the sample sintered from the plate-like precursor is 655 MPa,which is higher than that of the sample sintered from the flower-like precursor(524 MPa).Different reaction mechanisms and dispersing stabilities of the TiC particles at different temperature conditions should be accounted for the differences between the two samples.

Neon-concentration dependent retarding effect on the recrystallization of irradiated tungsten:Experimental analysis and molecular dynamics simulation

Recrystallization is a critical issue in ITER and future fusion devices as it strongly deteriorates the prop-erties and serving life of tungsten divertor armor.In this work,we investigated the influence of neon on the recrystallization behavior of tungsten and its underlying mechanism by both experimental anal-ysis and molecular dynamics simulation.Rolled tungsten was irradiated by neon ion with energies of 10 and 100 keV and fluences of 1×10^(18)-1×10^(20)m^(−2),followed by isochronal annealing at various temperatures(1473-2073 K)for 1 h.The results indicate that 10 keV neon ion irradiation slightly in-hibits tungsten recrystallization due to the shallow depth distribution of neon(35 nm),whereas 100 keV neon ion irradiation exerts a significant retarding effect on tungsten recrystallization when neon con-centration reaches∼102 appm.Grain boundaries can hardly move with neon concentration reaching 400 appm,above which the retarding effect is very strong and varies little.The retarded recrystallization is mainly attributed to the pinning effect of neon clusters/bubbles on the grain boundary movement dur-ing annealing.Large neon clusters are found to play a predominant role in the retarding effect and the un-clustered neon atoms show a very limited effect on the grain boundary movement.By comparing the neon effects on recrystallization in this work and helium effects reported in the literature,we found that helium exhibits a relatively stronger recrystallization retarding effect than neon.This is because helium is more easily to diffuse into grain boundaries and form larger clusters due to its high diffusivity and low self-trapping energy in tungsten,which can greatly impede the migration of grain boundaries.This work provides a fundamental insight into the neon-concentration dependent recrystallization retarding effect in tungsten and the corresponding underlying atomic-level mechanisms.

Microstructure characteristics of 12Cr ferritic/martensitic steels with various yttrium

12Cr ferritic/martensitic steels with 0, 0.1 wt%, 0.2 wt% and 0.3 wt% theoretical yttrium(Y) additions were fabricated by vacuum inducting melting and casting method. Solubilities of Y in the 12Cr steels are0.027, 0.078 and 0.17 for 12Cr-0.1 Y, 12Cr-0.2 Y and 12Cr-0.3 Y, respectively. Phase transformations and microstructure characteristics under different heat-treatment schedules were investigated. The starting temperature of ferrite-to-austenite transformation A^(c1) are maintained about 850℃, but the finishing temperature of ferrite-to-austenite transformation A^(c3) are about 950, 970, 980 and 1000℃ for 12Cr-0 Y,12Cr-0.1 Y, 12Cr-0.2 Y and 12Cr-0.3 Y, respectively, which indicates that A^(c3) increases gradually with the addition of Y. Martensite accompanied with a few δ-ferrite is the dominant structure in all the steels. The amount of δ-ferrite shows a strong dependence with the Y content and austenitizing temperature. Area fraction of δ-ferrite increases with the content of Y, which is the ferrite favouring element. The minimum amount of δ-ferrite are achieved at 950℃ for 12Cr-0 Y, 12Cr-0.1 Y, 12Cr-0.2 Y and 1000℃ for 12Cr-0.3 Y.Besides, more carbides precipitate along the martensite laths and grain boundaries in the Y-bearing steel due to the redistribution of carbon between austenite and ferrite resulting from the ferrite favouring element of Y.

First Results of Characterization of 9Cr-3WVTiTaN Low Activation Ferritic/Martensitic Steel

Ferritic/martensitic steels with Cr of 9%-12% （in mass percent） are favourable candidates for fuel cladding tube and in-core components of supercritical water-cooled reactor. 9Cr-3WVTiTaN low activation ferritic/martensitic steel, designated as China Nuclear Steel- I （CNS- I ）, was patterned after T91 steel （modified 9Cr-lMo） for the reactor. The idea of low activation material and microalloy technology was introduced into the design of the steel. The hardening, tempering and transformation behaviour of CNS- I steel was investigated. The steel has advantages in tensile properties at elevated temperature relative to zircaloy that has been widely used as cladding material for conventional light water reactors. CNS- I steel exhibits tensile properties and impact toughness comparable to T91 steel which exhibits availability in the present fission reactors and fast breeder reactor but includes undesired radioactive elements such as molybdenum and niobium.

Recrystallization temperature of tungsten with different deformation degrees

Pure tungsten （PW） and W-1 wt% La203 （WL10） were prepared by powder metallurgical route fol- lowed by the swaging ＋ rolling process. The logarithmic strains are 0, 0.37, 0.58, and 0.98 for WL10 and 0, 0.58 for PW. Heat treatments were performed at temperatures var- ied from 1,573 to 2,173 K to determine the recrystalliza- tion temperature. Recrystallization temperatures are 1,973 and 2,173 K for WL1 （logarithmic strain of 0.37） and WL3 （logarithmic strain of 0.98）, respectively. But in the case of WL2 （logarithmic strain of 0.58）, full recrystallization is not achieved at temperature of above 2,173 K. Further- more, the recrystallization temperature of PW with loga- rithmic strain of 0.58 is at least 300 K lower than that of the equivalent WLI0 sample. Moreover, the increase of recrystallization temperature inhibits the strength degra- dation of WL2： samples lose 4 % and 22 % strength when annealed at 1,573 and 1,973 K compared with room tem- perature （RT） sample. Finally, the texture evolution for the swaged ＋ rolled WL10 is significantly related to the deformation degree： the dominated orientation is （001） for WL2 while （110） for WL3.

Metallurgical characterization of Ti-bearing 9Cr low activation martensitic steel

Two heats of low activation martensitic （LAM） steels with Ti and Ta （denominated as 9Cr-Ti and 9Cr-Ta）, respectively, developed as candidate structure materials for nuclear reactor were characterized. This paper was focused on the effect of titanium on the microstructures and mechanical properties of 9Cr LAM steel in as-received condition （normalized at 950 ℃ for 30 min with water quenching plus tempered at 780 ℃ for 90 min with air cooling）. Chemical analysis and microstructure observation were conducted on 9Cr-Ti and 9Cr-Ta with optical microscopy, X-ray diffraction analysis, scanning electron microscopy and transmission electron microscopy. Impact properties and tensile strengths were measured with Charpy impact experiments and tensile tests. The results indicated that 9Cr-Ti and 9Cr-Ta were fully martensitic steels in as-received condition. MX type and M23C6 type precipitates were observed distributing along boundaries of prior austenite grains and martensite laths in 9Cr-Ti.The addition of titanium accelerated the precipitation of TiC and TiN, and produced much finer grains in 9Cr-Ti than 9Cr-Ta at the same normalization temperature. Mechanical properties tests showed the ductile brittle transition temperatures of 9Cr- Ti and 9Cr-Ta were about -90℃ and -85℃, respectively. The ultimate tensile strengths at room temperature and 600℃,were 680 MPa and 365 MPa for 9Cr-Ti, and 660 MPa and 335 MPa for 9Cr-Ta, respectively. The favorite impact toughness and tensile properties of 9Cr-Ti could be attributed to the fine grains in as-received condition.

Neutron irradiation effects on mechanical properties of ITER specification tungsten

In this contribution,we present the results of recent neutron irradiation campaign performed in the material test reactor BR2(Belgium)on pure tungsten.We have applied various irradiation conditions and sample geometry to assess the effect of neu-tron irradiation on hardness,bending,tensile and fracture mechanical properties.The investigated material is a commercially pure tungsten plate fabricated according to the international thermonuclear experimental reactor(ITER)specification for the application in the divertor plasma-facing components.The neutron irradiation covers a large span of temperatures and damage doses,ranging from 600 to 1200°C and 0.1-1 dpa.The obtained mechanical properties were analyzed to deduce the shift of the ductile to brittle transition temperature(DBTT)applying bending,tensile and fracture toughness-testing procedures.Then,a correlation of the fracture toughness with the change of the hardness was established.The obtained results are compared with the already published results on another ITER specification grade produced in the form of a rod.The presented and discussed results show that the performance of the compared grades in terms of the irradiation-induced embrittlement is similar,and that the irradiation in the high-temperature region(600-800°C)causes a considerable DBTT shift already at 0.2-0.5 dpa.

Effect of alloying on the properties of 9Cr low activation martensitic steels

Two types of 9Cr low activation martensitic steels （named 9Cr-1 and 9Cr-2） were developed in University of Science and Technology Beijing. 9Cr-1 and 9Cr-2 were produced by vacuum induction melting method, and examinations of the microstruc- tures were carried out with X-ray diffraction analysis, optical microscopy, scanning electron microscopy and transmission electron microscopy. The ultimate tensile strength and yield tensile strength were evaluated with tensile tests. The impact properties were characterized with Charpy impact experiments. The results indi- cated that 9Cr-1 and 9Cr-2 on as-received condition （95 ~C/30 min/water quenching plus 780 ~C/90 min/air cooling） were flflly martensitic steels free of ^-ferrite. The ul- timate tensile strength of 9Cr-1 and 9Cr-2 were 695 MPa and 680 MPa, respectively. However, 9Cr-2 showed a fine grain size of 4.8 pm, and its value of ductile-brittle transition temperature （DBTT） was -90 ~C under as-received condition. The additions of vanadium, titanium and boron accelerated the formation of MX precipitates and resulted in fine grains and precipitates. The fine grains effectively reduced the value of DBTT from -60 ℃ to -90 ℃ with identical upper shelf energy （USE）. The decrease in silicon concentration of 9Cr-2 induced a slight reduction in ultimate tensile strength from 695 MPa to 680 MPa.

Homogeneity analysis of Y-bearing 12Cr ferritic/martensitic steel fabricated by vacuum induction melting and casting

Advanced oxide metallurgy technique was adopted to produce 100-kg Y-bearing 12Cr ferritic/martensitic steel via vacuum induction melting and casting route. Subsequently, nine specimens at top, middle and bottom regions of the sheet were char-acterized to evaluate the homogeneity of chemical composition, microstructure and mechanical properties. The small vibra-tion of hardness (200–220 HBW), ultimate tensile strength (672–678 MPa), yield strength (468–480 MPa), total elongation (26.2%–30.5%) and Charpy energy at room temperature (98–133 J) and at ??40 ℃ (12–40 J) demonstrated that mechanical properties’ homogeneity of Y-bearing steel was acceptable although slight Y segregation and inhomogeneous microstructure occurred at the bottom. Furthermore, the effect of Y content on microstructure characteristics and mechanical properties was explained and the comparison of failure mechanism for the dual-phase steel between tensile test (i.e., quasi-static loading) and Charpy test (i.e., dynamic loading) was discussed in detail.

Preparation of a diamond coating by the CVD method on the tungsten substrate and its resistance to D plasma irradiation

After the surface of tungsten(W)alloys were roughened by laser,chemical vapor deposition(CVD)of diamond coatings were deposited on three tungsten substrates of pure W,W-1 wt.%La2O3 and W-0.5 wt.%TiC.Under the same growth parameters,the presence of the second phase in the tungsten matrix impeded the growth rate of diamonds,so a completed diamond coating with the thickness of 45μm and a grain diamond size of 10μm was obtained only on pure tungsten substrate after running for 10 h.Scanning electron microscopy,X-ray diffraction and Raman spectroscopy tests proved that the obtained diamond coating was compact with a high purity and regular morphology.To verify the chemical and structural stability,the as-obtained diamond-coated tungsten materials were exposed to an ion flux of 1.4×10^(21)ions m^(−2)s^(−1) in D plasma for 30 min.After irradiation,neither delamination,dramatic coating failure nor entire erosion of the coating(graphitization)was observed.The diamond coating can be an effective protective layer to stop tungsten atoms from splashing into the plasma.