Motokuni Eto

Low cycle fatigue properties of 8Cr–2WVTa ferritic steel at elevated temperatures

The effects of test temperature and tension holding on the fatigue properties of reduced activation ferritic/martensitic 8Cr–2WVTa (F-82H) steel were investigated by conducting low cycle fatigue tests at temperatures ranging from RT to 650°C under the axial strain-controlled condition with strains ranging from 0.5% to 2.0%. Fatigue life data were formulated as strain-life equations. A large reduction in the fatigue life was recognized at temperatures above 600°C. Softening without showing a saturated region was observed in the fatigue softening curves at temperatures above 600°C. Tension holding during fatigue tests reduced the fatigue life at 400°C, 500°C and 600°C. The microstructural examination showed that the large softening during cycle was associated with carbide (M23C6) coarsening and Laves phase (Fe2W) precipitation at 600°C.

Nanoindentation behavior of a two-dimensional carbon-carbon composite for nuclear applications

Abstract Nanoindentation tests on a carbon fiber and carbon matrix composite (C/C composite) for nuclear applications were carried out over a load range from 50 μN to 20 mN on two different crosssections, normal and parallel to the fiber axis. For reference, an isotropic nuclear graphite was also examined. Both the composite and graphite revealed a large amount of elastic recovery on the first loading-unloading indentation curves and inelastic hysteresis on the subsequent cycling curves. The analysis of mean indentation pressure data on the basis of the Weibull statistics indicates that the data on the C/C composite are separated into two groups with different Weibull moduli; the group of the lower modulus with higher mean pressure is attributed to the fiber, and the higher modulus with lower mean pressure to the matrix. The average of mean indentation pressure is obtained as 101 MPa for the fiber and 35.5 MPa for the matrix, while graphite has an intermediate value of 52.2 MPa. There is a slight dependence of mean stress pm on the applied load P in all the materials. However, the trend appears to be different between the two materials; pm decreases with increasing P for the C/C composite, while the reverse is true in graphite. Youngs moduli of the graphite and the two constituents of the C/C composite are estimated from the stiffness of the unloading indentation curves at 10.7 GPa for the graphite as 6.71 GPa for the fiber and 1.97 GPa for the matrix. The values estimated for the graphite and the fiber are comparable to those of the bulk graphite and C/C composite, respectively.

Evaluation of toughness degradation by small punch (SP) tests for neutron-irradiated 214Cr-1Mo steel

Mechanical properties correlations between the small punch (SP) test and conventional tensile, Charpy impact and fracture toughness tests were investigated on neutron irradiated 214Cr-1Mo ferritic steel. Estimation of radiation induced changes on tensile strength, elastic-plastic fracture toughness (JIC) and ductile-to-brittle transition temperature (DBTT) are thought to be basically possible by mechanical properties correlations when based on sufficient pre-irradiation data.

The effect of compressive prestressing on the mechanical properties of some nuclear graphites

Abstract The effects of prestressing in compression at room temperature on the Youngs modulus, the tensile strength, and the compressive strength of some nuclear graphites were examined. Also the effects of strain rate in the range of 10 −5 ~ 10 −2 sec −1 on the room temperature tensile and compressive strengths were investigated. The Youngs modulus and the tensile strength of the graphites decreased as the compressive prestress increased above a particular prestress level. However, the compressive strength of the graphites did not change appreciably for the above range of strain rates. The decrease in Youngs modulus and tensile strength with increasing prestress is considered to be due to two factors: an increase in the dislocation density and an increase in the size and number of cracks produced. It is inferred that the decrease in the strength at higher prestress levels is due to the formation and propagation of cracks with increasing prestress, since some growth of cracks was observed in the higher prestressed graphites.

High temperature young's modulus of a fine-grained nuclear graphite oxidized or prestressed to various levels

Abstract Effects of oxidation and compressive prestress on high temperature Youngs modulus were examined on a fine-grained isotropic graphite, IG-110. Oxidation was performed in air at 500°C up to a burn-off level of 10%. Prestress level was either 60% or 80% of the mean compressive strength. Both oxidation and prestressing caused decreases both in the room temperature and high temperature moduli, which were explained by a crack opening and growth model. An empirical equation by which the high temperature modulus of oxidized specimens can be estimated from the room temperature modulus of the unoxidized specimen was obtained as E (T, d) E(RT, d o ) = ƒ(T)( d d o ) n(T) . Here, E(T, d) and E(RT, do) are the high temperature modulus of the oxidized specimen with apparent density d, and the room temperature modulus of the unoxidized specimen with apparent density do, ƒ(T) is a polynomial in the third order of T, and n(T), a linear function of T. Annealing treatment or heat-up and cool-down process caused a partial recovery of the modulus of the oxidized specimen. This recovery was explained with the aid of a model: The internal stress generated as a result of the anisotropy of the thermal expansion of crystallites becomes smaller for the oxidized material because of the preferential oxidation of the binder region.

The effect of pre-stressing and annealing on the young's modulus of some nuclear graphites

Abstract Studies have been made on the decrease in the dynamic Youngs modulus of some nuclear graphites caused by pre-stressing or cyclic loading in compressive mode and its recovery due to annealing up to 2000 °C. It is found that the decrease in the modulus is well represented by the formulae, E E 0 = − Aσ 2 +B for ex-truded graphites and (E − E c ) (E 0 −E c ) = B exp ( − Aσ 2 ) for moulded ones, where E, Ec and E0 are the modulus at a pre-stress σ, that at pre-stress equal to fracture stress and that with no pre-stressing, respectively. A and B are constants. Annealing experiments show that the moduli of extruded graphites recover almost completely below 1000 °C, while in the case of moulded graphites complete recovery is not attained even at 2000 °C. The activation energy obtained for the recovery of an extruded graphite below 1000 °C is about 0.6 eV.

Mechanical properties and damage behavior of non-magnetic high manganese austenitic steels

Abstract Fe–Cr–Mn steels have been considered as materials of structural components for fusion reactor because of their low induced-radio-activity compared with SUS316 stainless steels. It has been expected to develop a non-magnetic steel with a high stability of the austenitic phase and a strong resistance to irradiation environments. For these objectives, a series of the Fe–Cr–Mn steels have been examined by tensile tests and simulation irradiation by electrons. The main alloying compositions of the steels developed are; C:0.02–0.2 wt%, Mn: 15 wt%, Cr: 15–16 wt%, N: 0.2 wt%. These steels were heat-treated at 1323 K for 1 h. The structure of the steels after the heat-treatment was austenite single phase. The yield stress of the steels was 350–450 MPa and the elongation were 55–60%. When the steels of high C and N was electron-irradiated at below 673 K, no voids were nucleated and only small dislocation loops were formed with high density. The austenite phase was also stable during irradiation below 673 K. Thus, newly developed high manganese steels have excellent mechanical proprieties and high irradiation resistance at relatively low temperature.

Quality evaluation of graphites and carbon/carbon composites during production of JT-60U plasma facing materials

Abstract The variations of physical and mechanical properties have been investigated for three grades of isotropic graphites and four grades of carbon/carbon (C/C) composites on the basis of the sample inspection data which have been obtained during production of the first wall and divertor plate materials for JAERI Tokamak-60 Upgrade (JT-60U). The evaluated properties are density, electrical resistivity, coefficient of thermal expansion (CTE), thermal conductivity, bending, tensile and compressive strengths. It is found that the maximum standard deviations normalized by the mean values are 22.7% for the C/C composites and 9.2% for the isotropic graphites. This scatter of the material quality should be considered in the design of the isotropic graphite and C/C composite armor tiles. Correlations between these properties are also observed for several materials.

Irradiation creep properties of a near-isotropic graphite

Abstract Two irradiation creep tests on near-isotropic graphite (SM1-24) for HTGRs were performed at around 900 °C in the JMTR. Neutron fluences ranged from 5.50 × 10 24 n/m 2 (E> 29 fJ) to 12.4 × 10 24 n/m 2 (E> 29 fJ) , depending on the position of the specimen. Irradiation creep strain (ϵ0) was obtained from the equation ϵ c = (σ/E 0 )[1-exp(−bΦ)] + KσΦ , by measuring dimensional changes in unloaded and loaded tensile specimens before and after irradiation, where E0 is the Youngs modulus before irradiation, K the creep coefficient, and b a constant. The value of K was estimated assuming that 1-exp(−bΦ) ∼-1 over the range of neutron fluence tested here. Mercury porosimetry was employed to add consideration to the mechanism of irradiation creep using unloaded and loaded specimens. The irradiation creep strain is proportional to stress and to neutron fluence for larger fluences. The irradiation creep coefficient is in inverse proportion to Youngs modulus before irradiation, KE 0 = 0.247 . From the values of the average Youngs moduli before irradiation for two irradiation creep tests, the creep coefficient was estimated to be 3.03 × 10 −29 (MPa/m 2 ) −1 and 3.18 × 10 −29 (MPa/m 2 ) −1 , respectively. The mercury pore diameter distribution changes upon irradiation, that is pores smaller than 10 μm disappear partly, the total porosity decreases, and the stress tends to facilitate disappearance of the pores. The Youngs modulus increases as a result of irradiation. The increase in Youngs modulus after a creep tests is smaller than that after irradiation only. The experimental result obtained here is consistent with the explanation for the mechanism of irradiation creep in which two to six interstitial clusters as a pinning point to basal slip disappear during the irradiation creep test.

Metal-hydride Characterization and Mechanical Properties of Ti-6A1-4V Alloy

Hydrogen absorption/dessociation properties and mechanical properties were examined in a Ti-6A1–4V alloy exposed to highly pressurized hydrogen gas of 0.1–5 MPa at elevated temperatures of 373 K to 773 K. Pressure-Concentration-Temperature (P-C-T) test results of the materials exhibit a large amount of hydrogen absorption up to 3 wt% H at 473 K with plateaus as observed in typical hydrogen absorption materials and the specimens tested are broken into pieces with the same level of hydrogen contents. Hydrogen dissociation from the material is not observed below 773 K. The results of bending tests and fractography show no degradation of mechanical properties of the materials with hydrogen contents of less than 0.1 wt% H. It is concluded from these test results that there is no mechanical degradation of Ti-6A1-4V vacuum vessel material for the tokamak device, JT-60SU due to metal-hydride reaction between Ti-6A1-4V and fuel hydrogen induced by hydrogen absorption mechanism into Ti-6A1-4V under the hydrogen pressure of 0.01–1 Pa H between room temperature and 773 K, in which JT-60SU is being designed, mainly because oxide layer formed on Ti-6A1-4V restrains metal-hydrogen reaction into the material.