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TEM observation and tensile test were examined for vanadium alloys irradiated in a temperature control rig in JMTR at 290 °C with damage level ranged from 0.003 to 0.06 dpa. With the increase of the neutron dose, irradiation hardening could be observed in all the vanadium alloys except for the V–5Nb alloy. In the case of pure vanadium, the relationship between irradiation hardening and neutron dose was described as Δ σ ∝ ( ϕt ) 0.35-0.53 . For V–5Cr alloy and V–4Cr–4Ti–0.1Si alloy, the dose dependence on irradiation hardening increase was shown as Δ σ ∝ ( ϕt ) 0.8 and Δ σ ∝ ( ϕt ) 0.8-1.0 , respectively. From the TEM observation, the hardening source of radiation-induced defects was mainly determined to be dislocation loops for pure vanadium, loops with voids for V–5Cr and, loops and {100} precipitates for V–4Cr–4Ti–0.1Si and V–3Fe–4Ti–0.1Si alloys. From the strain rate dependence of 8% stress for V–4Cr–4Ti–0.1Si alloys tested at RT, the strain rate sensitivity, m = 1/σ*( d σ/ d ln( d e/ d t)) shows positive. Therefore, the dynamic interaction between interstitial impurities and dislocation is not strong in V–4Cr–4Ti alloys in the temperature range from RT to 290 °C. A discrepancy of deformation mode of irradiated V–4Cr–4Ti–0.1Si alloys with 0.068 dpa could be seen when the charpy impact test indicated the brittle behavior and the tensile test indicated the ductile behavior at room temperature. It can be explained by the difference of strain rate for the value of yield stress between tensile test and charpy test and the critical fracture stress.

Vanadium alloys, including the highly purified V-4Cr-4Ti alloy, were irradiated in liquid lithium up to a damage level of 3.7 dpa in the HFIR at 425 °C and 598 °C. Neutron irradiation caused an increase of the ductile–brittle transition temperature (DBTT) and irradiation hardening was observed. Adding titanium to the V-Cr alloys was effective for increasing irradiation hardening at 425 °C. For highly purified (Zr-treated) V-4Cr-4Ti alloys the irradiation hardening was significantly reduced at both 425 °C and 598 °C. However, microstructural observations after the irradiation experiments showed that there was no significant difference in microstructure between the original and the highly purified specimens. It is suggested that the reduction of irradiation hardening in the highly purified V-4Cr-4Ti alloys was caused by the configuration and distribution of interstitial impurities in the neutron-irradiated specimen matrix. Controlling the impurities in V-4Cr-4Ti alloys has a very important effect for improving their mechanical properties that take place under neutron irradiation at around 400 °C.

A fusion reactor will yield a lot of high energy neutrons, and some of them will stream out of a plasma vacuum vessel and penetrate a blanket system and reach superconducting magnets which provide high magnetic field to confine high energy ionized particles. Under the neutron irradiation, the magnet materials will be activated and the properties will change. In this study, a research network on study of the irradiation effects is introduced and some data recently obtained are presented. By neutron irradiation, the critical current of the Nb3Sn wire increased and the critical field did not change. The organic insulation materials were degraded at neutron fluence of 1 × 1022 n/m2. In addition, the outline of an installation plan of 15.5 T superconducting magnet into radiation control area is described and study issues are explained. The project is undergoing as a program of Nuclear Basic Infrastructure Strategic Study Initiative.

The effect of specimen shape on the micro-crack growth behavior of F82H was investigated by the low cycle fatigue test at room temperature in air at the total strain range of 0.4–1.5% using three types of the miniature hourglass-type specimens with different stress concentration factors. In case of the high strain range conditions above 0.8%, almost no effect of the specimen shape on the fatigue life, micro-crack initiation life and the crack growth rate was observed. In case of the low strain range conditions below 0.6%, the effect of specimen shape on the fatigue life was observed. The fatigue life was considered to be dependent on the stress concentration factor. Based on the micro-crack growth analysis, the difference of the fatigue life under the low strain range conditions was attributed not to the crack growth rate but to the micro-crack initiation life dependent on the specimen shape.

Vanadium alloys, including the highly purified V-4Cr-4Ti alloy called NIFS-Heat2, in the form of sodium-enclosed irradiation capsules, were irradiated up to a damage level of 5 dpa in the Joyo reactor at temperatures from 395 to 601 °C. An increase of the ductile – brittle transition temperature (DBTT) by neutron irradiation and irradiation hardening were observed. No significant loss of ductility was observed even for irradiation at 395 and 450 °C. The addition of titanium to V-Cr alloys was effective for irradiation hardening at high temperature. Hydrogen uptake in the cleaning process during dismantling of the irradiation capsules caused ductility loss of the highly purified V-4Cr-4Ti alloys; the alloys recovered their ductility when they were annealed at 400 °C in vacuum. The uniform formation of Ti(OCN) precipitate was suppressed in highly purified V-4Cr-4Ti alloys irradiated in Joyo in a liquid-sodium environment.

In order to investigate the effect of the environment on the irradiation creep properties of highly purified V–4Cr–4Ti alloys, neutron irradiation experiments with sodium-enclosed irradiation capsules in Joyo and lithium-enclosed irradiation capsules in HFIR-17J were carried out using pressurized creep tubes (PCTs). It was found that the creep strain rate exhibited a linear relationship with the effective stress up to 150 MPa at 458 and 598 °C in the Joyo irradiation experiments. For HFIR-17J irradiation at 425 °C, the creep strain rate also exhibited a linear relationship with the effective stress up to 150 MPa. The activation energy of the irradiation creep and irradiation creep stress factor were estimated to be 46 kJ/mol K and 1–2, respectively. No significant difference in the irradiation creep behavior between liquid-sodium and liquid-lithium environments could be seen.

Ultra-fine grained (UFG) W-TiC consolidates are very promising for use as divertors in fusion reactors, however, the assurance of room-temperature ductility of UFG W-TiC remains unsettled. The assurance requires a sufficient degree of plastic working for the consolidates and thus overcoming of poor plastic workability in UFG W-TiC by applying superplasticity. Therefore, the magnitudes of elongation to fracture and flow stress which are important measures for plastic working were examined for UFG W-(0–1.5)%TiC (in wt%) at 1673–1973 K where superplasticity occurs without appreciable grain growth. It is shown that the elongation and flow stress are strongly dependent on TiC addition and atmosphere (Ar, H2) during mechanical alloying (MA). As the TiC addition increases, the elongation significantly increases without appreciable increase in the flow stress level. W-TiC fabricated with MA in H2 exhibits larger elongation and larger strain rate sensitivity of flow stress than W-TiC with MA in Ar. These results were applied to perform plastic working and the room-temperature bend test results for plastic worked W-1.0%TiC are shown.

In order to perform irradiation experiments in a liquid metal environment in a nuclear reactor, an irradiation technique with sodium bonding irradiation capsules was developed and a series of neutron irradiation experiments with sodium bonding irradiation capsules were performed in Joyo. The design and fabrication of sodium bonding capsules, sodium filling into capsules, capsule loading to Joyo, irradiation experiments, dismantling for irradiated capsules, removing the irradiated specimens from sodium-filled capsules, and sodium cleaning of the irradiated specimens were established through this study. Using the Joyo irradiation with the sodium bonding capsules where irradiation temperature was distributed uniformly, the irradiation creep experiment for highly purified V-4Cr-4Ti alloys, NIFS-Heat, was carried out and the knowledge about the irradiation creep behavior of the alloys was obtained.

Electrical conductivity of yttrium-doped perovskite-type barium-cerium oxide ceramics (BaCe 0.9 Y 0.1 O 3-δ ), implanted with 10 keV H 2 + ions, was measured in situ under fission reactor irradiation. An increment of the electrical conductivity, called radiation induced conductivity (RIC), was observed with increasing ionizing dose rate. The RIC for the specimen with implanted H at 1.1 kGy/s and irradiation temperatures 473-673 K was higher by about four orders of magnitude than the base conductivity without radiation at 0 Gy/s, and was about two orders of magnitude higher than that without H. The RIC is attributed to electronic excitation as well as hydrogen enhanced diffusion. The RIC greatly depended on the irradiation temperature, but was insensitive to the fast neutron fluence in the range 3.3-7.4 x 10 23 n/m 2 . The results show that the radiation induced defects, produced by neutron collisions, and radiolysis have no influence on the electronic and protonic conduction.

Small to ultra-small 1/3 size pre-cracked Charpy and 1.65 x 1.65 x 9 mm deformation and fracture minibeam (DFMB) specimens of the F82H IEA heat were irradiated to 0.02 and 0.12 dpa at 290 °C in the Japanese Materials Test Reactor. Nominal cleavage transition temperature shifts, based on the measured toughness, K jm (T), data (ΔT m ) as well as reference temperature shifts (ΔT 0 ) found after size-adjusting the K jm (T) data yielded ΔT m/0 ≈ 27 ± 10 and 44 ± 10 at the two doses, respectively. Using measured yield stress changes (Δσ y ), the C 0 = ΔT 0 /Δσ y = 0.58 ± 0.14 at 0.12 dpa, is in good agreement with data in the literature. The dynamic transition temperature shift, ΔT d , derived from DFMB tests, was ≈ 30 ± 20 °C at 0.1 dpa, also in good agreement with the estimated ΔT 0 shifts. The ΔT d and ΔT 0 are also in excellent agreement with a ΔT 0 = C 0 Δσ y (dpa, T i ) hardening-shift model, where the Δσ y (dpa, T i ) was found by fitting a large database of tensile properties.

The water cooled shield blanket made of Type 316L SS for the international thermonuclear experimental reactor (ITER) has potential issues related to stress corrosion cracking (SCC). Shroud mock-ups and boat samples taken from the core shroud of the boiling water reactor (BWR) with SCC were investigated from the viewpoint of microstructures and nano-hardness. Fine grains and deformation bands were observed in the hardened surface thin layers of the shroud mock-up, where hardness profiles in the ground portion was different from those in the milled portion. In the fine grain region, crevices were found only in the ground surface. In the core shroud, hardened surface regions were also found. Results showed that the crevices found on the ground surface could be one possible factor for SCC initiation.

In order to elucidate the effects of pre-irradiation microstructures and irradiation conditions on radiation embrittlement and radiation-induced ductilization (RIDU), fine-grained Mo–0.2 wt%TiC specimens with high and low reduction rates in plastic working, which are designated as MTC-02H and MTC-02L, respectively, were prepared by powder metallurgical methods. The specimens were neutron irradiated to 0.1–0.15 dpa with controlled 1-cycle and 4-cycle heating between 573 and 773 K, and 473 and 673 K, respectively, in JMTR. Vickers microhardness and three-point bending impact tests and TEM microstructural examinations were made. The degree of radiation embrittlement, assessed by DBTT shift due to irradiation, was strongly dependent on the reduction rate and cycle number. The 4-cycle irradiation suppressed the radiation embrittlement compared with the 1-cycle irradiation, and the suppression was much more significant in MTC-02L than in MTC-02H. The observed behavior is discussed in connection with RIDU and microstructural evolution caused by the 4-cycle irradiation.

To examine radiation effects on electrical insulating performance of candidate ceramic materials developed for the V/Li blanket system, in situ conductivity measurements for bulk specimens of Y2O3, CaZrO3 and Er2O3 were carried out during irradiation with fission neutrons in JMTR (Japan Material Testing Reactor). Evaluated radiation induced conductivities (RIC) of the specimens were 5.2 × 10−9 S/m for Y2O3 at ∼8 Gy/s (50 °C), 1.8 × 10−8 S/m for CaZrO3 at ∼40 Gy/s (35 °C) and 2.1 × 10−8 S/m for Er2O3 at ∼50 Gy/s (45 °C), respectively. The present results were consistent with the values extrapolated from the previous data obtained by DT neutron irradiations at low dose rate of up to ∼0.1 Gy/s. The flux dependence of the data indicates that RIC of the candidate ceramics with different radiation species could be well correlated using Gy/s as a parameter. The present results showed that the electrical degradation due to RIC will not prevent the achievement of required insulating performance for the blanket system.

Weld samples of fusion reactor vanadium alloy (NIFS-HEAT-2) were neutron-irradiated at 563 K up to 4:5 � 10 23 neutrons/m 2 (0.08 dpa). The recovery of irradiation hardening, degradation of impact properties, and damage structures in the weld metal were investigated after post-irradiation isothermal annealing at temperatures between 673 and 1073 K. Irradiation hardening and the decrease in impact absorbed energy at 77 K were larger for the weld metal than for the base metal. Recovery of the hardening by post-irradiation annealing was coincident with recovery of the impact absorbed energy in both the weld metal and the base metal. Recovery of the mechanical properties required postirradiation annealing at 1073 K for 1 h for the weld metal, which was 100 to 200 K higher than that for the base metal. The dislocation loops introduced by the neutron irradiation are likely to account for the hardening. The dislocation loops were observed even after annealing at 973 K in the weld metal, whereas they disappeared in the base metal. The characteristics of the radiation defects in the weld metal and the mechanisms for irradiation hardening and embrittlement are discussed.

A fabrication process for pressurized creep tubes (PCTs) for a highly purified V–4Cr–4Ti alloy, NIFS-heat2 was established. No increase in impurity contents in PCTs was detected during the tube manufacturing process. In a preliminary thermal creep test, homogeneous deformation was observed over entire tube length, which verifies reliability of creep measurement by using the present PCTs.

One key issue for small specimen test techniques is to clarify specimen size effects on test results. In consideration of size effects on determining the ductile-to-brittle transition temperature (DBTT) in Charpy impact testing, a method to evaluate the plastic constraint loss for differently sized Charpy V-notch (CVN) specimens is proposed and applied to a ferritic–martensitic steel, 2WFK, developed by JNC. In the method, a constraint factor, α , that is an index of the plastic constraint is defined as α=σ ∗ /σ y ∗ . Here, σ ∗ is the critical cleavage fracture stress which is a material constant and σ y ∗ is the uniaxial yield stress at the DBTT at the strain rate generated in the Charpy impact test. The procedures for evaluating each of σ ∗ and σ y ∗ are described and a result of σ ∗ and σ y ∗ , thus the value of α , is presented for different types of miniaturized and full-sized CVN specimens of 2WFK.

Weld specimens of the reference low activation vanadium alloy, NIFS-HEAT-2, were irradiated up to the neutron fluence of 4.5 × 10 23 nm −2 (0.08 dpa) at 563 K in JMTR. The weld metal showed larger irradiation hardening than that of the base metal, and the weld metal showed an embrittlement at 77 K. However, irradiation hardening at the weld metal was effectively reduced by post weld heat treatment at 873 and 1223 K. Mechanisms of irradiation hardening, and behavior of C, N and O impurities in the weld metal was discussed, based on the characterization of radiation defects.

Radiation induced changes of electrical conductivity for alumina varying density were investigated during fission reactor irradiation. Radiation induced conductivity (RIC) for dense and porous alumina increased due to electronic excitation with increasing the reactor power. The RIC at the reactor full power of 50 MW with ionization dose rate of 2.3 kGy/s and fast neutron flux of 1.4–1.6 × 1017 n/m2 s was higher by about two orders of magnitude than base conductivities at 0 MW. The RIC for porous alumina was about 3.3 times as much as that for dense alumina. The RIC of dense and porous alumina at 50 MW decreased at the initial fluence and reached a constant at fast neutron fluences of about 5.5 × 1022 and 2.5 × 1022 n/m2, respectively and subsequently kept up to the resultant ionization dose of 4.4 × 103 MGy and the fast neutron fluence of about 2.9 × 1023 n/m2.

Radiation effects in components and materials will be one of the most serious technological issues in nuclear fusion systems realizing burning-plasmas. Especially, diagnostic components, which should play a crucial role in controlling plasmas and understanding the physics of burning-plasmas, will be exposed to high-flux neutrons and gamma rays. Dynamic radiation effects will affect the performance of components substantially from the beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their service. High-power-density fission reactors will be the only realistic tools to simulate the radiation environments expected to occur in burning-plasma fusion machines such as the International Thermonuclear Experimental Reactor (ITER), at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibres, were irradiation-tested in a fission reactor, to evaluate their performances in heavy radiation environments. Results indicate that ITER-relevant radiation-resistant diagnostic components could be developed in time, although there are still some technological problems to be overcome.

W–5% Re–HfC alloys with aimed HfC contents of 0.5, 1 and 3 wt% were fabricated by powder metallurgy. Their thermal diffusivity and lattice constant were measured before and after neutron irradiation at 330 K up to a fast neutron (E>1 MeV) fluence of 1×1024 n/m2. Chemical analysis and lattice constant measurements revealed that Hf existed not as HfC but mainly as a solute in the alloys. The thermal diffusivity of the alloys with HfC less than 1% nominal content was higher than that of W–5% Re, specifically at higher temperatures, while the diffusivity of the alloy with 3% HfC (nominal content) was lower than that of W–5% Re. Only the alloy with 3% HfC showed a distinct decrease of the thermal diffusivity after neutron irradiation and its recovery after annealing. The lattice constants of the W–5% Re alloys increased with the increase of the Hf content. Only the lattice constant of the alloy with 3% HfC represented a considerable increase after neutron irradiation.

The reduced activation ferritic/martensitic steel, RAFs F82H IEA heat has been fatigue-tested at ambient temperature under diametral strain controlled conditions. In order to evaluate the effects of radiation damage and transmutation damage on fatigue characteristics, post-neutron irradiation and post-helium ion implantation fatigue tests were carried out. Fracture surfaces and fatigue crack initiation on the specimen surface were observed by SEM. Low-temperature irradiation caused an increase in stress amplitude and a reduction in fatigue lifetime corresponding to radiation hardening and loss of ductility. Neutron irradiated samples showed brittle fracture surface, and it was significant for large strain tests. On the other hand, helium implantation caused delay of cyclic softening. However, brittle crack initiation and propagation did not depend on the helium concentration profiles.

Oxyhydrate (OH) doped and fluorine (F) doped fused-silica core optical fibers were irradiated in a fission reactor at 370–380 K with a fast neutron flux of about 5×1016n/m2s and a gamma-ray dose rate of about 500 Gy/s for about 24 days. Oxyhydrate and fluorine dopings improved radiation resistance of optical fibers especially in the visible range. A fluorine doped and heat-treated optical fiber showed best radiation resistance and its radiation induced loss in a visible range is about 20 dB/m after it was irradiated up to a fast neutron fluence of 1×1023 n/m2. Results are implying that fused silica core optical fibers can be used nearer to a burning plasma for plasma diagnostics and remote sensing.

Magnetic probes for plasma diagnostics in the International Thermonuclear Experimental Reactor (ITER) were developed and tested in a fission reactor, the Japan Materials Testing Reactor. Performance of coils made of mineral insulating cables could be analyzed by a proposed equivalent electrical circuit but were disturbed by electrical drift, presumably caused by a phenomenon called radiation induced electromotive force. However, results were promising and development of an ITER-relevant magnetic probe is within view.

In order to clarify the effects of varying temperature during irradiation on the irradiation hardening of 9Cr–2W steels, tensile tests and positron annihilation lifetime measurements were carried out following the varying temperature irradiation (220/420°C and 340/530°C) utilizing a so called multi-section and multi-division controlled irradiation capsule in JMTR. After the irradiation at 220°C to 0.053 dpa, the steels show irradiation hardening as much as 110 MPa. The hardening was almost completely diminished immediately after the elevation of the irradiation temperature to 420°C. Subsequent irradiation at 420°C up to 0.14 dpa did not cause any hardening. The results of positron annihilation lifetime measurements indicate that microvoids are formed by the irradiation at 220°C but disappear upon elevating the temperature to 420°C and are then formed again by the subsequent irradiation at 420°C up to a total dose of 0.14 dpa. This behavior may be interpreted in terms of decomposition of interstitial loops or migration of small interstitial loops during temperature elevation. There is no good correlation between irradiation hardening and formation of microvoids in neutron-irradiated reduced-activation martensitic steels.

Mechanical property changes of Fe–XCr–2W–0.2V,Ta (X: 2.25–12) low activation ferritic/martensitic steels including Japanese Low Activation Ferritic/martensitic (JLF) steels and F82H after neutron irradiation were investigated with emphasis on Charpy impact property, tensile property and irradiation creep properties. Dose dependence of ductile-to-brittle transition temperature (DBTT) in JLF-1 (9Cr steel) irradiated at 646–700 K increased with irradiation up to 20 dpa and then decreased with further irradiation showing highest DBTT of 260 K at 20 dpa. F82H showed similar dose dependence in DBTT to JLF-1 with higher transition temperature than that of JLF-1 at the same displacement damage. Yield strength in JLF steels and F82H showed similar dose dependence to that of DBTT. Yield strength increased with irradiation up to 15–20 dpa and then decreased to saturate above about 40 dpa. Irradiation hardening in 7–9%Cr steels (JLF-1, JLF-3, F82H) were observed to be smaller than those in steels with 2.25%Cr (JLF-4) or 12%Cr (JLF-5). Dependences of creep strain on applied hoop stress and neutron fluence were measured to be 1.5 and 1, respectively. Temperature dependence of creep coefficient showed a maximum at about 700 K which was caused by irradiation induced void formation or irradiation enhanced creep deformation. Creep coefficient of F82H was larger than those of JLF steels above 750 K. This was considered to be caused by the differences in N and Ta concentration between F82H and JLF steels.

In order to clarify the mechanisms of weld cracking, especially for heat affected zone cracking in heavily neutron irradiated stainless steels and to establish a measure to evaluate crack susceptibility, a mini-sized Varestraint (variable restraint) test machine for hot laboratory operation was designed and fabricated. This unique PIE facility was successfully applied in the hot laboratory of IMR Oarai Branch of Tohoku University. The maximum restraint applied was 4% at the surface of the specimen. Specimen surface morphology and specimen microstructures were inspected by video microscope, SEM and TEM. Under the 2% surface restraint condition, clear formation of heat affected zone (HAZ) crack was observed for the case of neutron irradiation to produce 0.5 appm He and of 2.4 kJ heat input by TIG.

Miniature Charpy V-notch impact test specimens of commercial reactor pressure vessel (RPV) steels having high and low copper contents were irradiated at the different irradiation positions with neutron flux levels of ∼6 × 10 14 , ∼7 × 10 15 , and ∼8 × 10 16 n m −2 s −1 ( E > 1 MeV) to fluence levels ranging from ∼6 × 10 21 to ∼7 × 10 22 n m −2 ( E > 1 MeV) at temperatures of about 50°C to 150°C in the Japan Materials Testing Reactor (JMTR). The results showed that the radiation-induced increases in ductile-to-brittle transition temperature (ΔDBTT) at a neutron flux level of ∼6 × 10 14 n m −2 s −1 were greater than those for neutron flux level of ∼7 × 10 15 n m −2 s −1 . The neutron flux effect on embrittlement tended to be more pronounced in the lower neutron fluence range of ∼6 × 10 21 –∼1 × 10 22 n m −2 than in the higher fluence level of ∼7 × 10 22 n m −2 , and also to be larger for the low copper steel than for the high copper steel, although the ΔDBTT for the high copper steel was larger than that for the low copper steel regardless of neutron fluence or flux. The displacement dose rate effect identified by the data converted to the ΔDBTT for the full size Charpy specimens from those for the miniature Charpy specimens was consistent with that based on the comparison of the results in the literature.

The effect of neutron irradiation on Charpy impact properties of an ODS ferritic steel developed by PNC was studied. The miniaturized Charpy V-notch (MCVN) specimens (1.5 × 1.5 × 20 mm) of two orientations (longitudinal, called 1DS-L, and transverse, 1DS-T) were irradiated to fluence levels of (0.3–3.8) × 1026 n/m2 (En > 0.1 MeV) between 646 and 845 K in JOYO. MCVN specimens before and after the irradiation were subjected to instrumented Charpy impact tests. The test results and fracture surface observations showed that in the unirradiated state the steel showed no ductile-to-brittle transition behavior until 153 K regardless of orientation and the upper shelf energy of the steel was as high as that of a high-strength ferritic steel without dispersed oxide. Such excellent impact properties were essentially maintained after the irradiation although an appreciable decrease in absorbed energy occurred by higher temperature irradiations at and above 793 K.

Ductile–brittle transition (DBT) behavior of 9%Cr-2%W reduced-activation martensitic (RAM) steels has been investigated following neutron irradiation in the fast flux test facility, materials open test facility (FFTF/MOTA) at different temperatures. Both the irradiations at 663 and 733 K cause an increase in DBT temperature, while the irradiation at 663 K induces the hardening and the softening at 733 K. Microstructural observation by transmission electron microscope (TEM) revealed that small dislocation loops existed in the specimen irradiated at 663 K and no such a loop, but relatively large M6C carbides and Laves phase were formed by the irradiation at 733 K. There appears to be a linear dependence between ΔDBTT and ΔσY in neutron irradiated RAM steels when irradiation induces the hardening. Irradiation embrittlement accompanied by the softening is considered to be due to reduction of cleavage fracture stress caused by the irradiation-induced recovery of the martensitic structure, namely decrease in dislocation density and formation of large precipitates.

For applying to optical diagnostics for fusion reactor, two types of radiation-resistant optical fibers have been fabricated and their optical phenomena were examined in the JMTR. The optical absorption due to radiation-induced defects and formation of color centers, is concentrated in the wavelengths less than 700 nm. By contrast, the absorption was quite low in the wavelength range from 700 to 1500 nm. The optical fibers appeared to have good radiation resistance and survived neutron irradiation up to 2 × 10 24 n/m 2 . During JMTR irradiation, strong optical emission from 400 to 1400 nm with a sharp peak intensity at 1270 nm was observed. The peak value at 1270 nm is directly proportional to the reactor power. The results of absorption and fluorescence phenomena suggest the possibility of multi-parameter sensing for fusion reactor diagnostics.

Acquisition of systematic irradiation data are essential for understanding fundamental processes of irradiation effects and for establishment of a reliable database for irradiation effects in fusion reactor materials. It will take several years with expensive several different irradiation rigs in a fission reactor irradiation. There, it will take a long time to carry out the needed iterations between irradiation tests and evaluation and materials development. An irradiation rig was developed to carry out irradiation under multiple temperatures and irradiation fluences. Irradiation tests of fusion reactor materials were successfully carried out using the rig in Japan Materials Testing Reactor.

Several kinds of optical diagnostics are planned in a fusion reactor. Complicated optical systems such as periscopes are thought to be primary candidates for optical measurements, especially for visible wavelengths. However, optical fibers have several advantages over such optical systems. Also, the optical fibers could be a far better transmission line for signals under a high electromagnetic field. However, they have been considered vulnerable to heavy irradiation. In this study, several kinds of optical fibers were irradiated in the Japan Material Testing Reactor (JMTR). The optical transmissivity in fibers was measured in situ during fast neutron and gamma irradiation, up to doses of 2×1024 n m−2 and 5×109 Gy, respectively. The irradiation temperature ranged from 300 to 700 K. For pure ionizing irradiation environments, some methods for improving the radiation resistance of optical fibers were indicated. The results showed that effects of the irradiation associated with fast neutrons would be different from the effects of pure ionizing irradiation. Some fibers were found to withstand the heavy irradiation, especially in an infrared wavelength range.

Pure, OH-doped and F-doped silica core optical fibers were irradiated in a fission reactor at 400 ± 10 K using an electric heater at a reactor power greater than 10 MW (20% of the full power). The temperature was not controlled well at the early stage of the reactor startup, when the temperature was about 320-340 K. The optical fibers were irradiated with a fast neutron (E> 1 MeV) flux of 3.2 x 10 17 n/cm 2 s and a gamma dose rate of 3 x 10 3 Gy/s for 527 h. Optical transmission loss at 850 nm was measured in situ during irradiation, A prompt increase in optical transmission loss was observed as irradiation started, which was probably due to dynamic irradiation effects caused by short-lived and transient defects and is probably recoverable when irradiation ceases. After the prompt increase in optical transmission loss, a so-called radiation hardening was observed in fibers containing OH. Radiation hardening was also observed in 900 ppm OH-doped fiber at the second startup. The optical transmission loss increased linearly with irradiation dose, denoted as the accumulated loss, which we believe is due to irradiation-induced long-lived defects. Accumulated loss dominates radiation-induced optical transmission loss in a fission reactor irradiation.

A radiation induced electromotive force (RIEMF) was measured in mineral insulated cables (MI-cables) under fission reactor irradiation. A negative electromotive force of about 10 V was generated in the center lead by the reactor irradiation. The RIEMF increased with increasing irradiation dose with about − 14 V measured under the reactor full power irradiation at 573 K with a total dose of 1.1 × 1010 Gy ionizing irradiation and 1.1 × 1024 n/m2 fast neutron irradiation. The measured voltage of the RIEMF was found to depend linearly on the reciprocal temperature. The observed RIEMF was a so-called current driven source and the measured current was linearly dependent on the reactor power and was about 50 nA for the 0.7 m long MI-cables at 573 K. Under full power, irradiation conditions were 6 W/g of a gamma flux (6 × 103 Gy/s ionizing dose rate) for iron and 5 × 1017 n/m2 and 1.1 × 1018 n/m2 fast and thermal neutron fluxes, respectively.

The effect of neutron irradiation on the low temperature toughness of TiC-dispersed molybdenum alloys that were recently developed for fusion applications was studied. Miniaturized bend bar specimens, 1 by 1 by 20 mm, of both the alloys with TiC additions to 1.0 wt% and TZM were irradiated to 8 × 10 23 n/m 2 ( E n > 1 MeV) at controlled cyclic temperatures between 573 and 773 K in the JMTR and were tested by impact three-point bending. It was shown that under irradiation conditions the developed alloys exhibited much higher toughness than TZM and the toughness increased with increasing TiC content. In addition, the ductility of the alloy with 1.0 wt% TiC increased drastically by the irradiation, by about five times, regardless of the occurrence of significant irradiation hardening. This notable ductilization bX irradiation is the first found in materials and is discussed in connection with radiation enhanced precipitates.

Fusion relevant helium generation can be realized by a technique called ‘triple fission method’. This is essentially an injector-foil technique where the specimen is sandwiched with a pair of foils of fissionable material, e.g. uranium. Energetic alpha particles emitted in a small fraction of fission reactions are injected to adjacent specimens. This technique has been applied for the first time to the irradiation of a reduced activation ferritic steel developed for fusion applications. The results of the first experiment performed in the JMTR is reported in this paper. It is concluded that this technique is useful for simulating helium effects in ferritic steels where other simulation techniques are not easily applicable.

Irradiation response on the tensile properties of 9Cr2W steels has been investigated following FFTF/MOTA irradiations at temperatures between 646 and 873 K up to doses between 10 and 59 dpa. The largest irradiation hardening accompanied by the largest decrease in the elongation is observed for the specimens irradiated at 646 K at doses between 10 and 15 dpa. The irradiation hardening appears to saturate at a dose of around 10 dpa at the irradiation temperature. No hardening but softening was observed in the specimens irradiated at above 703 K to doses of 40 and 59 dpa. Microstructural observation by transmission electron microscope (TEM) revealed that the dislocation loops with the a〈100〉 type Burgers vector and small precipitates which were identified to be M6C type carbides existed after the irradiation at below 703 K. As for the void formation, the average size of voids increased with increasing irradiation temperature from 646 to 703 K. No voids were observed above 703 K. Irradiation softening was attributed to the enhanced recovery of martensitic structure under the irradiation. Post-irradiation annealing resulted in hardening by the annealing at 673 K and softening by the annealing at 873 K.

An experiment to measure radiation-induced electrical degradation (RIED) in a sapphire sample and in three MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260°C to a fluence of 3 × 1024 n/m 2 ( E > 1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m. Even though the results for the sapphire sample are somewhat ambiguous because of an unexplained offset current of about 0.6 μA substantial degradation was not observed in the sapphire: instead, radiation-induced conductivity (RIC) seemed to decrease slightly during the experiment. Substantial increase in leakage current, that increased with applied electric field, occurred in the MgO-insulated cables. This increased conductivity disappeared when the reactor was shut down and sample temperature returned to ambient. However, the physical degradation apparently remained in the material while the reactor was off because restarting the irradiation brought the conductivity back to its previous, degraded, reactor-on value. This effect is different from the RIED effect reported by Hodgson but is similar to previous results reported by Shikama et al. Considerable data were taken to determine the sample temperature and leakage currents during the irradiation.

The effects of irradiation on the performance of optical fibers with silica core were examined during irradiation in a fission reactor, JMTR. The fibers performed well in terms of their radiation characteristics up to the fast neutron fluence of 1.06 × 10 20 n/cm 2 and the gamma ray dose rate of 4.3 × 10 9 Gy at 460 K. Three kinds of optical radiation induced by the reactor irradiation were observed during irradiation. Origins of these optical radiations were studied. Broad optical radiations centered at 0.45 and 0.73 μm are thought to be Cherenkov radiation. The sharp optical radiation at 1.27 μm was thought to be generated in the silica core by the gamma-ray irradiations in the reactor.

Fracture properties of the reduced activation martensitic 9Cr-1WVTa and 9Cr-3WVTa steels were investigated by carrying out instrumented Charpy impact tests and tensile tests at temperatures below room temperature after irradiation in the Japan Materials Testing Reactor at 493 and 538 K. Modified 9Cr-1MoVNb steel was also examined for comparison. The irradiation-induced increase in ductile-to-brittle transition temperature was 53, 26 and 40 K for the 1 3 size Charpy specimens of 9Cr-1WVTa, 9Cr-3WVTa and 9Cr-1MoVNb steels, respectively, which resulted primarily from the irradiation-induced increase in yield stress. The cleavage fracture stress was 1820–1870 MPa for the three steels in unirradiated conditions, which was scarcely affected by irradiation. The deflections to the maximum load and to the brittle fracture initiation were decreased by irradiation. In the tensile test, quasi-cleavage fracture occurred at 77 K in both unirradiated and irradiated conditions. The cleavage fracture stress was 1320–1380 MPa for the tensile specimens of the three steels, which was about 1.4 times smaller than that for the Charpy specimens.

The accumulated results of radiation effects on materials have clearly shown that temperature history is very important for the evolution of microstructures induced by irradiation. The Japan Materials Testing Reactor has several preferable features for temperature-controlled irradiation. Two kinds of techniques were applied to control the irradiation temperature. The control of thermal insulation along with heating by the electrical heater, which is the other technique of the temperature control, permits us to control the temperature in the range of 360–1100 K, regardless of whether the reactor is in the operation or out of operation.

An electrical resistivity of polycrystal α-alumina was measured at about 680 K in a dynamic pumping condition in a fission reactor, JMTR. In a good vacuum, we observed smaller RIC (radiation-induced conductivity) than our previous results. We observed RIED (radiation-induced electrical degradation) -like behavior. The results suggested that RIED would take off faster at a higher ionizing dose rate. In the mean-time, the take-off occurred at about the same displacement damage of about 0.03–0.05 dpa in the range of fast neutron flux of (3.4–15.1) × 10 13 n/cm 2 s.

The current status of small-scale specimen technology in Charpy impact testing for ferritic steels is presented, with emphasis on the effect of the notch dimensions (notch depth, notch root radius and notch angle) on the upper shelf energy (USE) and ductile-to-brittle transition temperature (DBTT). The USE for miniaturized specimens, normalized by Bb2 or (Bb 3 2 (B is the specimen thickness, b the ligament size), is essentially independent of notch geometry and has a linear relationship with the USE of full size specimens, regardless of irradiation and alloy conditions. The DBTT of miniaturized specimens depends strongly on the notch dimensions; this dependence of the DBTT decreases as the DBTT of full size specimens increase due to neutron irradiation or thermal aging. These results may be useful in determining the USE and DBTT for full size specimens from those for miniaturized specimens.

Optical dosimetry, utilizing the radioluminescence, the Cerenkov radiation, and the radiation induced optical absorption, has attractive features. However, it has a serious setback, namely, the optical signal changes in the course of irradiation and complicated calibration would be needed. Also, the intensity of radioluminescence would depend complicatedly on energy and kind of incident ions and quanta. Here, optical dosimetry, which could measure the electronic excitation dose rate, the atomic displacement, and the thermal neutron flux, separately and in-situ, is proposed, by analyzing behaviors of radioluminescence peaks in the fused silica (SiO 2 ), the chromium doped alumina (Al 2 O 3 -Cr 2 O 3 ) and the Li 2 ZrO 3 ).

In order to develop the small specimen technology in Charpy impact testing, the effects of V-notch dimensions on the test results were investigated for miniaturized specimens of a ferritic steel, Japanese Ferrite/Martensite Dual Phase Steel (JFMS). The miniaturized Charpy specimens had four different square cross-sections of 3.3, 2, 1.5 and 1 mm, and each of them had a variety of V-notch dimensions (notch depth, notch root radius and notch angle). All of the specimens were subjected to Charpy impact tests between 93 and 373 K using a specially instrumented impact machine. The fracture surfaces of all tested specimens were examined by scanning electron microscopy

The tensile properties of reduced-activation martensitic 9Cr-IWVTa and 9Cr-3WVTa steels for fusion reactors were investigated over the temperature range from room temperature to 873 K after neutron irradiation in the Japan Materials Testing Reactor at 538 K to fast neutron fluences of 2×10 23 and 3×10 24 n/m 2 (E>1 MeV). A conventional 9Cr-1MoVNb steel was also examined for comparison. The irradiation caused an increase in yield stress and a decrease in total elongation. Irradiation-produced defect clusters could not be seen directly by transmission electron microscopy. The increase in yield stress caused by irradiation Δσ y was proportional to the 1/4 power of the neutron fluence

In-reactor measurements of the spontaneous and long-term change in electrical conductivity of a-alumina were carried out in the JMTR fission reactor. The so-called radiation-induced conductivity, RIC, was observed and evaluated qualitatively as well as quantitatively. A long-term increase in electrical conductivity was observed, which is thought to be due to the so-called radiation-induced electrical degradation, RIED

Post-irradiation microstructure observations were carried out for five 9%Cr-2%W low activation martensitic steels of which the chemical compositions of elements such as B, Mn, Zr and W were altered. Neutron irradiations were performed in FFTF/MOTA cycle 10 irradiations (10 dpa at 638 K, 30 dpa at 693, 793 and 873 K). Transmission electron microscope observations revealed that void swelling as high as 0.2% occurred in boron (30 wppm) doped 9%Cr-2%W steel irradiated to 40 dpa at 693 K with a void density of 4.7 × 1020 m−2. Some voids were observed in each steel irradiated at 638 K, but the swelling was negligible. A void size distribution diagram indicated that small bubbles less than 5 nm in diameter were observed in boron doped steels, while no such small bubbles were observed in a steel without an addition of boron. No voids were observed after irradiations at 793 and 873 K in any steels. An addition of 2% Mn caused a significant increase in the void density accompanied by a decrease in the void size. No chi-phase was observe in any steels after the irradiations, while Laves-phase was observed in specimens irradiated at 873 K to 33 dpa. Void shapes in these steels were determined to be dodecahedron and cube type.

The effects of irradiation on stress corrosion cracking (SCC) and intergranular corrosion (IGC) susceptibility were investigated in solution-treated Fe19Cr9NiMn alloys and JPCA irradiated to 5.3×1024 n/m2 (E > 1 MeV) at 573 K. In Fe19Cr9NiMn alloys, the irradiation enhanced IGC i n boiling HNO3 + Cr6+ solution when the alloys contained phosphorus and silicon and induced SCC in all the alloys with strain rate tensile tests in 571 K water containing 32 ppm oxygen. With increasing phosphorus and silicon contents. IGC was promoted but IGSCC was suppressed after irradiation. The results indicated that these elements are not the main contributors to irradiation-assisted SCC, although they affect SCC behavior. The Japanese Prime Candidate Alloy (JPCA) had better SCC resistance than Fe19Cr9NiMn alloys under the present irradiation condition.