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Gamma radiolysis experiments on solutions of a mixture of sodium chloride (NaCl) and sodium bromide (NaBr) were conducted to confirm the validity of radiolysis calculations for simulated seawater s...

Radiolysis calculations of simulated seawater were conducted using reported data on chemical yields and chemical reaction sets to predict the effects of seawater constituents on water radiolysis. H...

We investigated the effect of radiolysis on pit initiation of zircaloy-2 in water containing sea salt. Changes in the composition of water containing sea salt were analyzed as well, both before and after gamma-ray irradiation. The characteristics of the resultant oxide films formed on zircaloy-2 were evaluated by X-ray photoelectron spectroscopy and electrochemical impedance spectroscopy. The experimental results showed that the pitting potential under irradiation was slightly higher than that under conditions in which no radiation was present, and that the pitting potential decreased with increasing chloride concentration in the presence as well as the absence of radiation. Solution analyses for water containing sea salt showed that hydrogen peroxide was generated by irradiation. The oxide film was composed of zirconium oxide and was made thicker during the irradiation. Under gamma-ray irradiation, the zircaloy-2 surface with an oxide film formed by radiolysis products was found to be resistant to pitting in the presence of chloride., 著者所属: 日本原子力研究開発機構(JAEA)

Cracking growth tests were conducted in high-temperature water containing hydrogen peroxide (H2O2) at 561–423 K to evaluate the effects of H2O2 on stress corrosion cracking (SCC) of stainless steel (SS) at temperature lower than the boiling water reactor (BWR) operating temperature. Small compact tension (CT) specimens were prepared from thermally sensitized type 304 SS. Despite the observation of only a small portion intergranular SCC (IGSCC) near the side groove of the CT specimen at 561 K in high-temperature water containing 100 ppb H2O2, the IGSCC area expanded to the central region of the CT specimens at 423 and 453 K. Effects of H2O2 on SCC appeared intensely at temperature lower than the BWR operating temperature because of a reduction in the thermal decomposition of H2O2. To estimate the environment in the cracks, outer oxide distribution on the fracture surface and the fatigue pre-crack were examined by laser Raman spectroscopy and thermal equilibrium calculation was performed.

Crack growth tests were performed in high-temperature water containing hydrogen peroxide (H 2 O 2 ) to evaluate the relationships between the crevice structure and H 2 O 2 on stress corrosion cracking (SCC) growth morphology of stainless steel (SS). Small compact tension (CT) specimens were prepared from thermally sensitized type 304 SS. 20–300 ppb H 2 O 2 was injected into the high-temperature water at 561 K. Intergranular SCC (IGSCC) and transgranular SCC were observed near the side grooves and the central region of the original CT specimens, respectively. Chevron notches were removed from the CT specimens after fatigue pre-crack introduction. Owing to pre-crack shortening, the IGSCC area expanded to the central region of the CT specimens and increased with H 2 O 2 concentration. The effects of H 2 O 2 on SCC appeared intensely near the surfaces exposed to high levels of H 2 O 2 . Microanalysis and distribution examination of oxide layers were performed and the percentage of H 2 O 2 remaining in the crack was calculated.

In-pile loop experiments are one of the key technologies that can provide an understanding of corrosion behaviors of structural materials in nuclear power plants (NPPs). The experiments should be s...

It is known that the presence of even a small amount of impurity in interstitial positions can, depending on temperature, have a drastic influence on the one-dimensional (1-D) motion of sel...

From the mid-seventeenth century, with the formation of Osaka city, members of the hinin status group, an organization originally composed of beggars and indigents alienated from all forms of ownership, became established in urban society within an organization called the “kaito fraternity of the four places.” Over time, members of Osaka’s hinin fraternity secured the right to beg as a means of survival and were entrusted with the duty of policing and providing relief to the “new hinin” and “wild hinin” who emerged on the margins of the hinin status group. As an extension of those activities, the hinin fraternity also came to perform a range of official police duties under the authority of the City Magistrate’s Office. While members of the hinin fraternity possessed specialized begging rights and official duties, those rights and duties existed inside a broader network of social relationships. Namely, members of the fraternity were only able to survive by begging because they maintained relationships with city neighborhoods and townspeople that provided alms. Similarly, members of the hinin fraternity were only able to perform official duties because they maintained a relationship with the magistrate’s office, which ordered them to perform those duties, and with the individual neighborhoods that employed “hinin watchmen.” In this paper, I focus on how the will of Osaka’s townspeople restricted efforts by members of the hinin fraternity to redistribute begging rights during the nineteenth-century. By doing so, this paper highlights the stratified and composite nature of early modern Japan’s status society.

To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 10 26 n/m 2 ( E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03–1.0 × 10 26 n/m 2 . Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

To examine the effects of minor elements on stress corrosion cracking (SCC) susceptibility of low carbon stainless steels with a work hardened layer, a high purity type 304 stainless steel was fabricated and minor elements, Si, S, P, C or Ti, were added. A work hardened layer was introduced by shaving on the surface of stainless steels. The specimens were exposed to a boiling 42% MgCl 2 solution for 20 h and the number and the length of initiated cracks were examined. SCC susceptibility of the specimen with P was the highest and that of the specimen with C was the lowest in all specimens. By magnetic force microscope examination, a magnetic phase expected to be a martensitic phase was detected near the surface. Since corrosion resistance of martensite is lower than that of austenite, the minor elements additions would affect SCC susceptibility through the amount of the transformed martensite.

Aluminum-doped type 316L SS (316L/Al) has been developed for the purpose of suppressing the degradation of corrosion resistance induced by irradiation in austenitic stainless steels (SSs). The electrochemical corrosion properties of this material were estimated after Ni-ion irradiation at a temperature range from 330 °C to 550 °C. When irradiated at 550 °C up to 12 dpa, 316L/Al showed high corrosion resistance in the vicinity of grain boundaries (GBs) and in grains, while severe GB etching and local corrosion in grains were observed in irradiated 316L and 316 SS. It is supposed that aluminum enrichment was enhanced by high-temperature irradiation at GBs and in grains, to compensate for lost corrosion resistance induced by chromium depletion.

Reduced activation ferritic steel is one of the leading structural material candidates for a nuclear fusion reactor, in which tungsten is introduced in order to improve the radiation resistance. Since the solute impurities have an effect on the embrittlement through segregation under irradiation, the stability of impurity elements should be elaborated. In the present study the segregation characteristics of tungsten and some general solute impurities in bcc iron were investigated nonempirically by first principles calculations, where the equilibrium segregation was considered via a regular solution model and the change in enthalpy for segregation were directly evaluated for comparison. Subsequently the energetic stabilities of impurity–impurity and impurity–vacancy pair were evaluated. The segregation enthalpy is influenced by the electronic interaction between the d-electron of Fe and the outer electron of the impurity element, and molybdenum and tungsten tend to prevent solute atoms from segregation.

The authors previously reported that magnetic flux leakage d ata, measured using a flux gate (FG) sensor, showed a correlation with the irradiation-assisted stress corros ion cracking (IASCC) susceptibility of neutron-irradiate d austenitic stainless alloys. This paper presents a study conducted to develop a diagnostic system that can detect IASCC precursors in stainless steels by measuring the magnetic properties of the material. The eddy current method and alternating current (AC) magnetization method were used, as these will be more practical for use in actual reactors. Probes were developed for these measurement methods, providing sufficient tolerance for en vironments in nuclear reactors. An attempt was also made to improve spatial resolution by manufacturing a smaller probe. A sensor system was designed for remote control, performance tests were conducted by measuring neutron-irradiated specimens and mock specimens, and magnetic data were evaluated by comparing the IASCC susceptibility of the specimens. It was proved that the sensor system developed in this study is capable of detecting IASCC precursors. Further developments necessary for application in actual nuclear reactors and the mechanism of correlation between magnetic properties and IASCC susceptibility were also discussed. Many of the problems experienced in existing light water reactors (LWR) are caused by damage to structural materials. Stress corrosion cracking (SCC) and irradiation-assisted SCC (IASCC) of structural materials have been especially serious problems, and it has not been possible to predict these in the reactor design process. These problems will become more significant when existing reactors start to age from now on, and when advanced reactors are operated and structural materials are used in more severe irradiation environments in future. It is therefore import ant to develop diagnostic techniques that can

The size- and spacing- dependent obstacle strength due to the Cu precipitation in α-Fe is investigated by atomistic simulations, in which the effect on phase transformation of Cu precipitation is considered by a conventional self-guided molecular dynamics (SGMD) method that has an advantage to enhance the conformational sampling efficiency in MD simulations. A sequence of molecular statics simulations of the interaction between a pure edge dislocation and spherical Cu precipitation are performed to investigate the obstacle strength associated with phase transformation. It was shown that the SGMD method can accelerate calculating the bcc to 9R structure transformation of a small precipitate, enabling the transformation without introducing any excess vacancies. Such metallographic structures increase the obstacle strength through strong pinning effects as a result of the complicated atomic rearrangement within the Cu precipitation.

Crack initiation and crack growth processes of irradiation assisted stress corrosion cracking of stainless steels were studied by slow strain rate testing (SSRT) in oxygenated high temperature 561 K water. In-situ observation was carried out during SSRT for type 304 stainless steel irradiated to a neutron fluence of 1.0 × 1025 n/m2 (E > 1 MeV) at 323 K in the Japan material testing reactor. The specimens were subjected to solution annealing, thermal sensitization, or cold working prior to neutron irradiation. The solution annealed material exhibited a combination of transgranular stress corrosion cracking (TGSCC) and ductile fracture, and almost all intergranular stress corrosion crackings were observed in the thermally-sensitized material. In the cold-worked material, cracking was introduced before the maximum stress was reached, and the fracture mode changed from TGSCC to ductile fracture to transgranular cracking together with the progress of crack growth in one direction.

In order to examine the stress corrosion cracking (SCC) susceptibility of reduced-activation ferritic/martensitic steel, F82H, slow-strain-rate-test (SSRT) was performed at various temperatures in oxygenated or hydrogenated water. Test specimens of F82H were heat-treated at various temperature conditions, or were cold-worked to simulate radiation hardening and machined to make single edge notch, or were neutron-irradiated at 493 K to 3.4 dpa. It was found that in unirradiated specimen, IGSCC occurred when specimen was normalized only, and TGSCC occurred when cold-worked (over 23%) and notched specimen was tested by SSRT at 573 K in oxygenated water. In irradiated specimen, TGSCC occurred, when SSRT was conducted at 573 K in hydrogenated (DH = 1 ppm) water or when the notched specimen was tested by SSRT at 573K in oxygenated (DO = 10 ppm) water.

Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension-type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 × 1025 n/m2 under a pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests are discussed and compared with the results obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC. From results of in-core and ex-core tests using pre-irradiated specime...

The two-dimensional intergranular stress corrosion cracking (IGSCC) growth model has been developed to simulate branching cracks of IGSCC. In the model, the IGSCC is grown using the "grain-scaled" factors such as the length and strength of grain boundary and so on. Especially, the corrosion of grain boundary and the influence of shear stress acting on the grain boundary are introduced in the model. Using the model, computer simulation of crack growth was carried out under several load conditions with changing the ratio of axial to shear stress against the grain boundary. As a result of the simulations, we found out that the cause of crack branching was the influence of shear stress against the grain boundary, and that the synergistic effect of shear stress and corrosion of grain boundary leads to the oblique crack growth.

In core structural materials of next generation reactors such as a liquid-metal cooled fast breeding reactor and a supercritical-water cooled thermal or first reactor, materials' degradation behavior by neutron irradiation damage and thermal (cyclic) stress should be considered with fair accuracy in design process (including maintenance and repair plans), because the materials are used under higher temperature gradients and higher neutron flux fields than those in the present light water reactors. In the current experiential design rules, service lives of core structural components were determined by the materials degradation such as the increase of ductile-to-brittle transition temperature after post irradiation examination data. However, other materials degradations such as irradiation-assisted stress corrosion cracking (IASCC), which occurs by the degradation synergistically interacting with radiation hardening, local chemical composition change, swelling and radiation creep, should be considered reasonably in the design process of the next generation reactors, because of the anticipation of the beneficial effects by synergy of radiation damage. The radiation hardening and local chemical composition change at grain boundaries due to radiation-induced segregation increased with increasing dose. Above some threshold dose, swelling increased rapidly with increasing dose. Residual stress due to thermal stress and welding procedure decreased with increasing dose. To predict material failure by IASCC with reasonable accuracy, in this study, each material degradation phenomenon with different dose dependence was modeled with consideration of radiation induced stress relaxation. And then the models were integrated to simulate the failure behavior for the duration of reactor operation period. In this paper, the models obtained by ion-irradiation experiments and compared by data from neutron irradiation experiments were presented, and the concept of our new evaluation method and the programming code for the failure simulation were outlined.

Effects of water temperature on the stress corrosion cracking (SCC) susceptibility were studied on a solid hot-isostatic pressed (solid-HIPed) type 316LN-ITER grade (316LN-IG) stainless steel in unirradiated or irradiated condition. Neutron irradiation was conducted at a nominal temperature of 473 K to a dose level of 1 dpa. Slow strain rate testing (SSRT) was performed over a temperature range from 423 to 603 K in oxygenated (DO = 10 ppm) or hydrogenated (DH = 1 ppm) water. In the unirradiated condition, intergranular (IG) type SCC occurred at 603 K in the oxygenated and in the hydrogenated water. In the irradiated condition, IASCC occurred above 573 K in oxygenated water. In the hydrogenated water, IASCC occurred at 513 K, but was suppressed at 573 K.

To examine the crack initiation and growth of irradiation assisted stress corrosion cracking (IASCC), in situ observations of the specimen surface were conducted during slow strain rate testing (SSRT) in oxygenated high purity water at 561 K. Each specimen of type 304 stainless steel was treated by solution annealing (SA), thermal sensitization (TS) or cold working (CW). After neutron irradiation of 1.0 × 1026 n/m2 (E > 1 MeV), the gage length section of specimen was observed through a window equipped in the autoclave of SSRT machine and images were recorded automatically at regular intervals. Crack initiation was observed immediately after the maximum stress during SSRT of each specimen. The ranking of mean crack growth rate based on the images was CW > TS > SA. The relative susceptibility to IASCC estimated by fraction of intergranular stress cracking on fracture surface was observe to be SA > TS > CW.

It is important to evaluate the effect of irradiation on the corrosion behavior of materials to be used in spallation neutron sources. Solution annealed high purity Fe–18Cr–12Ni specimens were used in this study. Ni3+ and H+ or He2+ ions were injected at 473–773 K. After corrosion test, the specimens were examined with atomic force microscope (AFM) to evaluate the corrosion behavior. It was shown that the corrosion rate of the irradiated area increased with increasing dose and temperature. H implantation accelerated corrosion. On the other hand, He implantation seemed to suppress corrosion. Mechanisms for these effects of the different irradiation conditions on the corrosion behavior are discussed.

To study the effects of minor elements on irradiation assisted stress corrosion cracking (IASCC), high purity type 304 and 316 stainless steels (SSs) were fabricated and minor elements, Si or C were added. After neutron irradiation to 3.5 × 10 25 n/m 2 ( E >1 MeV), slow strain rate tests (SSRTs) of irradiated specimens were conducted in oxygenated high purity water at 561 K. Specimen fractured surfaces were examined using a scanning electron microscope (SEM) after the SSRTs. The fraction of intergranular stress corrosion cracking (IGSCC) on the fractured surface after the SSRTs increased with neutron fluence. In high purity SS with added C, the fraction of IGSCC was the smallest in the all SSs, although irradiation hardening level was the largest of all the SSs. Addition of C suppressed the susceptibility to IGSCC.

Effects of water and irradiation temperatures on irradiation-assisted stress corrosion cracking (IASCC) of type 316 stainless steel were investigated. Type 316 stainless steel was irradiated at 333–673 K to a dose level of 16 dpa. Susceptibility to IASCC was evaluated by slow strain rate testing in oxygenated water in the temperature range of 513–573 K. Irradiation at 603 and 673 K caused IASCC in 513 K water, but irradiation below 473 K did not induce IASCC at 513 K. Specimens irradiated at 333 K did not show IASCC susceptibility in 513 K water, but high susceptibility was observed in 573 K water. Effect of irradiation temperature is discussed from the view points of microstructural and microcompositional changes.

The present work investigated the magnetic transition temperatures of four model alloys for simulating radiation induced segregation in austenitic stainless steels SUS 304. SQUID magnetometer was used to measure magnetization curves at temperature from 400 to 5 K. The results show that for model alloys with lower Ni content (8 and 10 wt%), there occurred spontaneous martensitic transformation and produced a ferromagnetic phase during the cooling process in SQUID magnetometer. For the model alloys with higher Ni concentration (20 and 30 wt%), no martensitic transformation was observed. Martensitic transformation starting temperature, Curie temperature and Neel temperature have been calculated for these alloys. It reveals that magnetic parameters are very sensitive to the change of chemical composition and microstructure.

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

Solid hot-isostatic pressed (solid-HIPed) joint specimens and those with or without thermal-cycled specimens of 316LN-IG were irradiated at about 473 K to 1 dpa in the high flux isotope reactor. Slow strain rate tests were conducted in a high-purity, oxygenated ( dissolved oxygen =10 wt ppm) water at 423, 513 and 573 K with strain rates of (2–10)×10−7 s−1. Tensile tests were also conducted in vacuum at the same temperature range. Fracture surfaces were observed by scanning electron microscopy. No specimen showed irradiation-assisted stress corrosion cracking (IASCC) susceptibility at 423 and 513 K in water. At 573 K, however, intergranular cracks were observed to form in HIPed specimens. It was concluded that the effect of HIPing to IASCC susceptibility is small.

Type 316L(N) stainless steel of the international thermonuclear experimental reactor grade (316L(N)-IG SS) is being considered for the first wall/blanket module. Hot isostatic pressing (HIP) technique is expected for the fabrication of the module. To evaluate the integrity and susceptibility to stress corrosion cracking (SCC) of HIPed 316L(N)-IG SS, tensile tests in vacuum and slow strain rate tests in high temperature water were performed. Specimen with the HIPed joint had similar tensile properties to specimens of 316L(N)-IG SS, and did not show susceptibility to SCC in oxygenated water at 423 K. Thermally sensitized specimen was low susceptible to SCC even in the creviced condition. It is concluded that the tensile properties of HIPed SS are as high as those of the base alloy and the HIP process caused no deleterious effects.

Stress relaxation of tensile type specimens under fast neutron irradiation at 288 °C has been studied for 316L stainless steel (SS) in the Japan Materials Testing Reactor. In-pile stress-relaxation tests were carried out at fast neutron fluence levels of 1.3×10 24 , 5.5×10 24 and 1.5×10 25 n/m 2 ( E >1 MeV). These tests were carried out at the applied total strain levels of 0.06%, 0.1%, 0.3% and 0.75%. In order to evaluate the thermal stress-relaxation behavior and to distinguish it from the irradiation induced stress-relaxation behavior, out-of-pile stress-relaxation tests were also performed at 288 °C in air using an electric furnace. This paper describes results of in-pile and out-of-pile stress-relaxation tests on 316L SS tensile specimens. These results are compared with the literature data by Foster et al. [J. Nucl. Mater. 252 (1998) 89] which were mainly obtained from bend beam specimens. Moreover, these experimental results are compared with analytical results obtained using Nagakawa's model [J. Nucl. Mater. 212–215 (1994) 541].

In the Japan Atomic Energy Research Institute (JAERI), in-pile strain measurement techniques have been developed using the Japan Materials Testing Reactor (JMTR). In order to evaluate the performance of fiber optic grating sensors under irradiation environment, heat-up and performance tests at elevated temperatures before irradiation and in-pile tests were performed in JMTR. It was determined that it is possible to measure strain under irradiation environment below 1×1023 n m−2 (E>1 MeV) by a fiber optic grating sensor, because in-pile temperature characteristics were in good agreement with out-of-pile test results.

In Japan Atomic Energy Research Institute, in-pile strain measurement techniques have been developing by using JMTR. In order to evaluate the performance of capsule type strain gage and fiber optic grating sensor under irradiation environment, the heat-up tests in electric furnace before irradiation and in-pile tests were performed. As for capsule type strain gage, it is found that gage factor almost did not change and both its electric resistance and strain output decreased by irradiation effect. A correlation was found between the change ratio of the electric resistance of the capsule type strain gage and fast neutron fluence. As for fiber optic grating sensor, it is possible to measure strain under irradiation environment below 1×1023n/mm2(E>1MeV) by this sensor, because in-pile temperature characteristic was in agreement with out-of-pile test results.

A material performance database for nuclear applications, which was named the JAERI Material Performance Database (JMPD), has been developed since 1986 with a view to utilizing various kinds of characteristic data of nuclear materials efficiently. The data stored in the JMPD are mainly fatigue crack growth data on low alloy steels, creep data on superalloys, tensile data on aluminum alloys and stress corrosion cracking data (Slow Strain Rate Testing (SSRT), crack growth rate, etc.) on austenitic stainless steels. Irradiation Assisted Stress Corrosion Cracking (IASCC) of austenitic stainless steels in high temperature water has been considered as a degradation phenomenon potential not only in light water reactors (LWRs) but in the rather common systems where the materials are exposed to radiation in the presence of water. This paper describes the present status of the JMPD, which is partially available through the Internet. Furthermore, some trials for the utilization of the system focused on the issues re...

A precipitation hardening type 630 stainless steel was irradiated in the Japan Materials Testing Reactor (JMTR) in contact with the reactor primary coolant. The temperature of the irradiated specimens was about 330 K. The fast neutron (E > 1 MeV) fluence for the specimens ranged from 1024 to 1026 m−2. Tension tests and fracture toughness tests were carried out at room temperature, while Charpy impact tests were done at temperatures of 273–453 K. Tensile strength data showed a peak of 1600 MPa at around 7 × 1024 m−2, then gradually decreased to about 1500 MPa at 1.2 × 1026 m−2. The elongation decreased with irradiation from 12% for unirradiated material to 6% at 1.2 × 1026 m−2. The fractography after the tension test revealed that the fracture was ductile. Fracture toughness decreased to about a half of the value for unirradiated material with irradiation. The cleavage fracture was dominant on the fractured surface. Charpy impact tests showed an increase of ductile–brittle transition temperature (DBTT) by 60 K with irradiation.

Solution-annealed, high-purity type 316 stainless steel and its heats doped with C, Ti, Si, P and S alone or together were irradiated at 513 K to a dose level of 1 dpa in the Japan Research Reactor No. 3 at Japan Atomic Energy Research Institute. After irradiation, transmission electron microscopy was carried out. In all alloys, Frank loops were mainly developed. Addition of Mo decreased the number density and the average diameter of Frank loops. Addition of C made the density increase and the diameter decrease, while addition of Si made the density and the diameter decrease.

The IASCC is one of the major concerns on the integrity of in-core materials not only for the LWRs but also for the materials to be cooled by water in fusion reactor. It is desired to irradiate test materials in a high temperature water environment in order to investigate simultaneous effects of irradiation and corrosive environments, which are essential for the study of IASCC. In the JMTR, an irradiation rig named the SATCAP was developed and applied for the IASCC study. Inside of this capsule, cooling water is boiling during irradiation due to heat generated by gamma heating, thus the temperature of the specimens is kept nearly at saturation temperature of water. Recently a new hybrid type SATCAP with electric heaters and vacuum control systems was developed to improve temperature control capability of the original SATCAP. A design and results of performance test of the new hybrid type SATCAP are reported.

Effects of irradiation temperature on the irradiation assisted stress corrosion cracking (IASCC) were studied on a high-purity type 304 model stainless steel (HP) and its alloy doped with carbon (HP + C). Solution-annealed and cold-worked specimens of the alloys were irradiated at 513 and 323 K up to 7-9 × 10 24 n/m 2 ( E > 1 MeV) and then susceptibilities to IASCC were examined by the slow strain rate testing (SSRT) in high-purity 573 K water. Solution-annealed specimens of both alloys, HP and HP + C, irradiated at 323 K showed a susceptibility to intergranular stress corrosion cracking (IGSCC). The specimen of alloy HP + C failed at a shoulder portion with very high susceptibility to IGSCC. Addition of carbon affected the fracture morphology and increased radiation hardening by irradiation at 513 K, while it seems to increase uniform elongation of specimens irradiated at 323 K. The 50% cold-worked specimens of both alloys showed no susceptibility to IASCC.

The austenitic stainless steels of type 304 and type 316 and the precipitation hardened type 630 stainless steel were irradiated to doses ranging from 5 × 1024 to 1.2 × 1026 at 323 K in the Japan Materials Testing reactor (JMTR). Tensile and fracture toughness tests have been carried out at 293 K. Type 304 and 316 steels in the solution annealed condition hardened by irradiation to about 700 MPa. Fracture toughness values for these steels decreased with dose. Only a small hardening occurred after irradiation for cold worked type 304 and precipitation hardened type 630 steels. The effect of irradiation on the fracture surfaces was small.

A low impurity Fe 18Cr 12Ni (HP) and its heats doped with Si and C (HP + Si and HP + C) were irradiated to 6.7 × 10 24 n/m 2 ( E ≫ 1 MeV) at 513 K. The slow strain rate tensile (SSRT) tests were carried out at a constant strain rate of 1.7 × 10 −7 s −1 in high purity, 573 K water. Scanning electron microscopy on the fracture surface revealed that HP and HP + Si failed mainly by the intergranular stress corrosion cracking (SCC), while the major failure mode in HP + C was the transgranular SCC. All alloys exhibited radiation hardening. HP + Si exhibiting the smallest hardening showed uniform elongation of 17%, while HP and HP + C did not. Transmission electron microscopy was also carried out. Frank loops and unidentified small clusters were formed in HP and HP + C, while only small clusters were observed in HP + Si.

Le statut de « bourgeois » (chōnin) dans le Japon de l’epoque premoderne est le produit de la construction, au tournant des xvie et xviie siecles, d’une multitude de villes nouvelles ou se concentra la population des marchands et des artisans. Selon les principes d’organisation sociale et politique de cette epoque, le statut de bourgeois garantissait aux citadins un droit de propriete fonciere en echange de l’accomplissement d’un « service » (yaku) envers l’autorite publique seigneuriale, impose au bourgeois dans le cadre de communautes autonomes de rue ou de quartier. A partir de l’exemple des deux grandes metropoles shogounales d’Edo et d’Ōsaka au xviie siecle, l’article se penche sur l’evolution de quelques rues, en montrant les changements rapides qui affecterent le statut de bourgeois, du fait des modifications de la structure de la propriete dues aux transactions foncieres, mais aussi de la montee de la population locative.

Susceptibility to the irradiation-assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR was evaluated through slow strain rate tensile (SSRT) test in oxygenated high purity water. SSRT tests were carried out at 300°C under controlled potential condition. The material had been cold-worked 20% and irradiated to 40 displacement per atom (dpa) at 425°C. Some of the specimens were thermally aged after irradiation at 900°C for 0.1 min, 30 min and 12 h to investigate the effect of aging on IASCC behavior. An applied potential enhanced stress corrosion cracking, strength and elongation of the material obtained by SSRT testing showed prompt recovery by thermal aging. SEM observation of fracture surfaces revealed marked change of dominant cracking mode from intergranular (IG) type to transgranular (TG) type.