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Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal-hydraulic analysis were coupled into an integrated code system. In the system, each code...

9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 °C, which surpassed that of heat-resistant austenitic steels although creep rupture strength of tempered martensitic steels is generally lower than that of austenitic steels at high temperatures. The measured creep rupture strength of 9Cr-ODS steel claddings at 1000 °C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 °C. This superior strength seemed to be owing to transformation of the matrix from the α-phase to the γ-phase. The transient burst strengths for 9Cr-ODS steel were much higher than those for 11Cr-ferritic/martensitic steel (PNC-FMS). Cumulative damage fraction analyses suggested that the life fraction rule can be used for the rupture life prediction of 9Cr-ODS steel and PNC-FMS claddings in the transient and accidental events with a certain accuracy.

The Nuclear Energy Agency (NEA) Expert Group on Innovative Structural Materials (EGISM) was established in 2008 under the guidance of the Nuclear Science Committee (NSC). Its objectives are to conduct joint and comparative studies to support the development, selection and characterisation of innovative structural materials that can be implemented in advanced nuclear fuel cycles, under long service lifetime and extreme conditions, such as high temperature, high dose/dose rate and corrosive chemical environments. In this context of growing interest and initiatives, the EGISM initiated at the beginning of 2018 an activity among its members to: • Identify, in a non-exhaustive way, the currently existing programs on innovative materials and fabrication processes in NEA member countries and China; • Establish a first cartography of the activities that are underway on these topics and identify common subjects and thematic; • Propose a Technology Readiness Level scale to estimate the maturity of both innovative materials and fabrication processes; • Carry out a reflection on what the enablers are to quickly climb this TRL scale, as well as the obstacles, in order to identify solutions to overcome them. This paper first gives definitions shared between the EGISM members on what are considered as advanced structural materials solutions. Next, some international initiatives for the accelerated development of high performance materials are presented both in non-nuclear and nuclear fields. Then, the methodology adopted for technology readiness assessment is explained. A non-exhaustive synthesis of the projects identified among the EGISM members on innovative structural materials and advanced manufacturing solutions such as additive manufacturing is presented. The TRL level of these projects is evaluated. A focus is also made on some of these projects to illustrate and explain the TRLs chosen as well as highlighting enablers or obstacles identified to climbing up the TRL scale.

The objective of this study was to investigate the effect of nitrogen concentration on mechanical properties and nano-structure of 9Cr oxide dispersion strengthened (ODS) ferritic/martensitic steel. 9Cr-ODS specimens with the wide range of nitrogen concentration, from 0.004 to 0.110 wt%, were systematically investigated by hardness and tensile tests and several microstructural characterization methods. Hardness and tensile strength at 973 K were significantly decreased as nitrogen concentration increased, due to the decrease in the amount of the residual α-ferrite phase. Coarse inclusions containing Y and Ti, which could negatively affect creep strength and processability, were formed, and that suggested degradation of the nano-particle distribution. The technical knowledge obtained in this study will contribute towards the setting of a reasonable nitrogen concentration specification for 9Cr-ODS steel. Keywords: ODS, Nitrogen, Ferrite, Martensite, α to γ reverse transformation, HT-XRD

A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in sodium bulk flow was set at 0.07 wt ppm in the calculation., 高温流動Na中での燃料被覆管からのCr溶出挙動に及ぼす各種試験パラメータ(Na中溶存Cr濃度、試験温度、軸方向温度勾配、Na流速)の影響を系統的に把握するための計算モデルを構築し、数値計算を行った。Cr溶出挙動が液体Na中の溶存Cr濃度の微量の変化に大きく影響される計算結果を得た。Na中のCr濃度は、Na冷却高速炉(SFR)内における燃料被覆管からのCr溶出挙動の予測・制御のため、注目すべき重要なパラメータであると考えられる。Na中溶存Cr濃度を0.07wtppmmとした場合の本モデル計算で得たODS鋼被覆管肉厚方向のCr濃度プロファイルは、BOR-60照射試験データとよく一致した。

We developed a non-contact deformation measurement system to accurately evaluate strain for post irradiation tensile testing, since conventional strain gages cannot be used for small size specimens. The strain calculated from cross-head displacement generally includes deformation from specimen shoulders, fixtures, and the test frame in addition to the deformation from the specimen gauge section. In our system, the distance between painted marks within the specimen gauge section was measured using a high resolution video camera to evaluate the specimen deformation during room temperature tensile testing. The test materials were F82H and 9Cr ODS steels irradiated up to ≈71 displacements per atom (dpa) at about 573 K in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Our system yielded accurate stress strain curves without deformations other than the specimen gage section, and the elongation was less than that calculated from cross-head displacement. This system can contribute to expanding the technically reliable database for the design activity of fusion reactor blanket, including the effects of irradiation on tensile properties. Keyword: Reduced activation ferritic/martensitic steel, Oxide dispersion strengthened steel, Post irradiation experiment, Small specimen technology, Tensile test, Strain

The aim of this study was to evaluate the tensile properties and microstructures of dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging at temperatures between 400 and 600 °C up to 30,000 h. Characterization of microstructure was carried out by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Microstructural analysis showed that the microstructure in the weld metals (WMs) consisted of lath martensite containing a small amount of residual austenite. Thermal aging hardening of WMs occurred at 400 and 450 °C due to the effects of both α-α’ phase separation and G-phase precipitation. However, there was no significant change in the total elongation, and fracture surfaces indicated that very fine dimpled rupture was predominant rather than the cleavage rupture. It was suggested that lath martensite phases enhanced the tensile strength due to phase separation, while residual austenite played a role in keeping elongation as a soft phase.

Oxide dispersion strengthened (ODS) steel cladding tubes have been developed for fast reactors. 9 chromium ODS and 11Cr-ODS tempered martensitic steels are prioritized for the candidate material in research being carried out at JAEA. In this work, fundamental immersion tests and electro-chemical tests of 9 to 12Cr-ODS steels were systematically conducted in various nitric acid solutions at 95$^{\circ}$C. The corrosion rate exponentially decreased with effective solute chromium concentration (Cr$_{eff}$) and nitric acid concentration. Addition of oxidizing ions also suppressed the corrosion rate. According to polarization curves and surface observations in this work, the combination of low Cr$_{eff}$ and dilute nitric acid could not prevent the active dissolution at the beginning of immersion, and the corrosion rate was high. In comparison, higher Cr$_{eff}$, concentrated nitric acid and addition of oxidizing ions helped to prevent the active dissolution, and suppressed the corrosion rate., 高速炉用に酸化物分散強化型(ODS)鋼燃料被覆管の開発が進められている。原子力機構では9および11Cr-ODS焼戻しマルテンサイト鋼をその候補材料としている。被覆管からの腐食生成物が再処理工程に及ぼす影響を見積もるためには、その腐食挙動を評価しておく必要がある。本研究では95$^{\circ}$Cの硝酸溶液中で基礎的な浸漬試験および電気化学試験を系統的に実施した。腐食速度は有効Cr濃度(Cr$_{\rm eff}$)および硝酸濃度の増加に対して、指数関数的に低下した。酸化性イオンの添加も腐食速度を低下させた。取得した分極曲線や浸漬後の表面観察結果より、低Cr$_{\rm eff}$と希硝酸の組み合わせでは浸漬初期に活性溶解が発生し、腐食速度が大きくなることが分かった。一方、高Cr$_{\rm eff}$と比較的高濃度の硝酸、あるいは酸化性イオンの添加の組み合わせは不働態化を促進し、腐食速度を低下させることが分かった。

The growth of oxide particles in FeCrAl-oxide dispersion strengthened steel (ODSS) considering an accident condition of the light-water reactor at above 1500 K was studied by using a high-temperature annealing. Oxide particles grew from 9 nm to more than 50 nm as maximum at 1623 K for 27 h, with decreasing their number density in two orders of magnitude. Most of the oxide particles in 15Cr-7Al were identified as YAM or YAP, while the oxide particles in 15Cr-7Al-0.4Zr were identified trigonal Y4Zr3O12. Zr addition to 15Cr-7Al ODSS accelerated the growth of the oxide particles, which is quite contrary to the effect of Zr addition during sintering as suggested in the literature. The kinetics of coarsening was characterized by an equation of Ostwald ripening. The diffusion activation energies obtained in the present materials were quite larger than the conventional diffusion activation energy of Y in alpha-iron. Gibbs free energy of oxides should be considered to discuss the coarsening. (C) 2017 Elsevier B.V. All rights reserved.

Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900–1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

Microstructure of 9Cr oxide dispersion strengthened (ODS) tempered martensitic steel (TMS) cladding and 12Cr-ODS ferritic steel cladding after very long-term internal pressurized creep tests was observed. The creep test conditions were 700°C for 46,504 h and 750°C for 45,550 h for 9Cr-ODS TMS and 700°C for 30,116 h for 12Cr-ODS ferritic steel. The stability of martensite and ferrite matrix structure, including carbide distribution, dislocation density, and lath boundary, and that of nano-particle distribution were investigated. It was confirmed that the nano-particles were stably present during the high-temperature and very long-term stress loading environment, contributing to the prevention of matrix structure degradation. The results of observation are consistent with the fact that the 9Cr-ODS TMS and 12Cr-ODS ferritic steel exhibit excellent creep strength and show no reduction in the slope of the stress versus time to rupture curve over a significantly extended period of time in the creep test.

The effect of thermo-mechanical treatments (TMTs) on the evolution of nano-structure in an oxide dispersion strengthened (ODS) ferritic/martensitic steel (Fe-9Cr-2W-0.22Ti-0.36Y2O3) was investigated. TMTs involve hot extruding and subsequent forging, which are expected to be part of a future industrial-scale manufacturing process of the ODS steel. It was shown that the ODS steel was composed of two phases — a fine-grained residual ferrite phase and a transformable martensite phase. The number density of the nano-sized particles in the residual ferrite phase was significantly higher than that in the martensite phase. The TMTs did not significantly affect the number density, but slightly affected the size distribution of the nano-sized particles in both ferrite phase and martensite phase. Moreover, the volume fraction of the residual ferrite phase decreased after TMTs. In summary, the TMT conditions could be a parameter which affects the nano-structure of the ODS steel.

The relationship among tensile strength, Vickers hardness and dislocation density for two types of 11Cr-ferritic/martensitic steel (PNC-FMS) was investigated after aging at temperatures between 400 and 800$^{\circ}$C up to 45,000 h and after neutron irradiation. A correlation between tensile strength and Vickers hardness was expressed empirically. The linear relationship for PNC-FMS wrapper material was observed between yield stress and the square of dislocation density at RT and aging temperature according to Bailey-Hirsch relation. Therefore, it was clarified that the correlation among dislocation density, tensile strength and Vickers hardness to aging temperature to aging temperature was in good agreement. On the other hand, the relationship between tensile strength ratio when materials were tested at aging temperature and Larson-Miller parameter was also in excellent agreement with aging data between 400 and 700$^{\circ}$C. It was suggested that this correlation could use quantitatively for separately evaluating irradiation effects from neutron irradiation data containing both irradiation and aging effects.

Oxide Dispersion Strengthened (ODS) ferritic/martensitic steels have an excellent high temperature strength primarily due to a dislocation pinning effect of nanometric oxide particles. In the present work, the interaction between oxide particles and dislocations in 9CrODS ferritic steel was investigated by both static TEM observation and in-situ TEM observation under dynamic straining conditions. The primary concerns of those observations were the obstacle strength of oxide particles and the type of interactions: attractive or repulsive. In the static observation, the majority (∼90%) of all interaction geometries was characterized as repulsive type. In the in-situ straining experiments, the obstacle strength α of oxide particles was estimated to be no greater than 0.80. The experimentally-determined obstacle strength is smaller than that of Orowan type impenetrable obstacle, whereas those oxide particles are, in theory, ideally strong obstacles. The gap between predicted and measured obstacle strength is attributable to cross-slip motion of screw dislocations on the oxide particles.

Materials for core components of fusion reactors and fast reactors, such as blankets and fuel cladding tubes, must be excellent in high temperature strength and irradiation resistance because they will be exposed to high heat flux and heavy neutron irradiation. Oxide dispersion strengthened (ODS) steels have been developing as the candidate material. Japan Atomic Energy Agency (JAEA) have been developing 9 and 11 Chromium (Cr) ODS steels for advanced fast reactor cladding tubes. The JAEA 11Cr-ODS steels were rolled in order to evaluate their anisotropy. Tensile tests and creep tests of them were carried out at 700 $^{\circ}$C in longitudinal and transverse orientation. The anisotropy of tensile strength was negligible, though that of creep strength was distinct. The observation results and chemical composition analysis suggested that the cause of the anisotropy in creep strength was prior powder boundary including Ti-rich precipitates., 核融合炉のブランケットおよび高速炉の燃料被覆管といった炉内機器の材料は、高熱流束と中性子重照射にさらされるため、高温強度と耐照射性に優れている必要がある。その候補材料として酸化物分散強化型(ODS)鋼の開発が進められている。原子力機構(JAEA)では先進高速炉の燃料被覆管用に9Crおよび11Cr-ODS鋼の開発を進めている。本研究ではJAEA-11Cr-ODS鋼を圧延し、その異方性を評価するため、圧延方向と横断方向について引張試験とクリープ試験を700$^{\circ}$Cで実施した。その結果、引張強さでは異方性を示さなかったが、クリープ強度では明瞭な異方性を示した。各種観察と元素分析の結果、クリープ強度異方性はTi析出物を内包した旧粉末境界が原因であると分かった。

Aiming at further improvement of high-temperature oxidation and corrosion resistance, 11CrODS steel with martensitic base structure has been previously developed, as a candidate fuel cladding material for 4th generation advanced nuclear reactors. In this study, the microstructure of 11CrODS steel was characterized by means of EBSD and nanoindentation hardness measurement. The continuous cooling transformation (CCT) diagram was constructed. Upper critical cooling rate, which is minimum cooling rate necessary to form martensitic structure, was derived to be 60 °C/min (3600 °C/h). In contrast, lower critical cooling rate preventing from martensite formation, was derived to be 10 °C/min (600 °C/h). An area fraction of so called residual ferrite was estimated by image processing of EBSD-IQ map to be 21% of the total area. This fraction of the residual ferrite in 11CrODS steel was evaluated by considering the driving force for α to γ reverse transformation.

This study evaluates the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins with austenitic stainless steel cladding were obtained from steady state irradiation in the Experimental Breeder Reactor II. Post irradiation examinations revealed that fission gas pressure was the main cause of creep damage in cladding, and that the stress contribution from fuel-cladding-mechanical-interaction was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to fission gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. On the other hand, the CDFs derived from thermal control creep equation showed the value less than that from the in-reactor creep rupture equation. These results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects, i.e. the in-reactor creep rupture equation, are used.

Accepted: 2015-07-29, 資料番号: SA1150179000

Oxide dispersion strengthened (ODS) steels are notable advanced alloys with durability to a high-temperature and high-dose neutron irradiation environment because of their good swelling resistance and mechanical properties under neutron irradiation. 9–12Cr-ODS martensite steels have been developed in the Japan Atomic Energy Agency as the primary candidate material for the fast reactor fuel cladding tubes. They would also be good candidates for the fusion reactor blanket material which is exposed to high-dose neutron irradiation. In this work, modification of the manufacturing process of 11Cr-ODS steel was carried out to improve its impact property. Two types of 11Cr-ODS steels were manufactured: pre-mix and full pre-alloy ODS steels. Miniature Charpy impact tests and metallurgical observations were carried out on these steels. The impact properties of full pre-alloy ODS steels were shown to be superior to those of pre-mix ODS steels. It was demonstrated that the full pre-alloy process noticeably improved the microstructure homogeneity (i.e. reduction of inclusions and pores).

Increasing Cr content from 9 to 12 mass% leads to superior corrosion and high-temperature oxidation resistances, and usually changes microstructure from martensite to a ferrite. To make transformable martensitic type of 12CrODS steels that have superior processing capability by using alpha/gamma phase transformation, alloy design was conducted through varying nickel content. The structure of 12CrODS steels was successfully modified from full ferrite to a transformable martensite-base matrix containing ferrite. This ferrite consists of both equilibrium ferrite and a metastable residual ferrite. It was shown that the fraction of the equilibrium ferrite is predictable by computed phase diagram and formation of the residual ferrite was successfully evaluated through pinning of alpha/gamma interfacial boundaries by oxide particles.

The dissimilar butt welding joint of 11Cr-ferritic/martensitic steel (PNC-FMS) and Type 316 austenitic stainless steel (SUS316) produced by electron beam (EB) welding was studied. This study was carried out to investigate optimization of EB welding and postweld heat treatment (PWHT) for the wrapper tube materials. Optimum EB welding conditions were a focus position of 30–40 mm and a welding speed of 1750–2000 mm/min, and optimum PWHT was performed after welding at 690 °C for 60 min. As a result, no formation of δ-ferrite was observed adjacent to the fusion zone, and the mechanical properties of the welds were similar to those of the base material. In this regard, EB welding is a proper fusion welding process for dissimilar PNC-FMS and SUS316.

A tensile test was performed at a loading direction perpendicular to elongated cold-rolled grains, and it confirmed the step of the scratched lines across the grain boundaries on the specimen surface, providing evidence for grain boundary sliding in oxide dispersion strengthened (ODS) steels. Dynamic recovery within the grains was also observed, and a simple model was constructed that consisted of grain boundary sliding and mismatch accommodation induced by vacancy flow. It was confirmed that such grain boundary sliding was suppressed in ODS steel relative to iron as a result of the pinning of the dislocation movement by the dispersed oxide particles.

An oxide dispersion strengthened (ODS) austenitic stainless steel was developed by mechanical alloying (MA) of advanced SUS316 stainless steel. A nano-characterization was performed to understand details of the effect of minor alloying elements in the distribution of dispersoids. It is shown that Y2O3 particles dissolve into the austenitic matrix after the MA for 6 h. Annealing at 1073 K or higher temperatures result in a distribution of fine oxide particles in the recrystallized grains in the ODS austenitic stainless steel. Additions of Hafnium or Zirconium led to the distribution of finer oxide particles than in samples without these elements, resulting in an increase in the hardness of the samples. The most effective concentration of Hf and Zr to increase the hardness was 0.6 and 0.2–0.3 wt%, respectively.

The microstructure and tensile properties at 973 K in 9CrODS steels were investigated with respect to various fractions of the residual ferrite from zero to 47 vol.%. The formation of the residual ferrite was discussed from a balance between a chemical driving force for α to γ reverse transformation and the oxide particle pinning force, while 0.04 mass% carbon sample could contain an equilibrium δ-ferrite. With increasing volume fraction of the residual ferrite, strength at 973 K increases but ductility decreases, which is attributed to the fact that the residual ferrite is harder than the tempered martensite.

The oxide dispersion strengthened (ODS) ferritic steels are expected as the structural materials for the advanced fusion reactors due to superior radiation resistance and high-temperature strength. Mechanically alloyed powders were consolidated by hot-extrusion and formed by hot-forging. Three types of ODS steels were manufactured: 9CrODS containing 16 % and zero % of the residual ferrite and 12CrODS containing 41 % of residual ferrite. Miniaturized specimens were prepared in the longitudinal direction parallel to finely elongated residual ferrite. The upper shelf absorbed energies (USE) vary in 41 J, 94 J and 134 J in full-size, which increase with increasing a residual ferrite fraction. Higher USE is attributed to larger crack propagation energy, which is induced by different fracture modes in the residual ferrite and tempered martensite.

In this study, identification of the crystal structure and analysis of the orientation relationship of oxide particles in an oxide dispersion strengthened austenitic stainless steel was carried out. High resolution transmission electron microscopy (HRTEM) and energy dispersive spectroscopy showed that most of the oxide particles had a faceted shape and consisted of a complex oxide, the anion-deficient fluorite structure Y2Hf2O7. Selected area diffraction patterns and HRTEM indicated that the faceted oxide particle has a cube-on-cube orientation relationship with the surrounding matrix. In addition, strain fields were observed around the oxide particle with given reflection conditions, indicating that it surrounds the oxide particle. The observed strain fields would affect glide dislocation pinning and the migration of irradiation-induced point defects.

11-12Cr oxide dispersion strengthened (ODS) tempered martensitic steels underwent manufacturing tests and their ferritic–martensitic duplex structures were quantitatively evaluated by three methods: high-temperature X-ray diffraction (XRD), electron probe microanalyzer (EPMA), and metallography. It was demonstrated that excessive formation of residual-α ferrite, due to increasing Cr content, could be suppressed by appropriately controlling the concentration of the ferrite-forming and austenite-forming elements on the basis of the parameter “chemical driving force of α to γ reverse transformation. 11Cr-ODS steel containing a small portion of residual-α ferrite was successfully manufactured. In the as-received condition, this 11Cr-ODS steel was shown to have satisfactory creep strength and ductility, both as high as those of the 9Cr-ODS steel, while its 0.2% proof strength at 973 K was lower than in the 9Cr-ODS steel.

Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel. Two types of ODS (9Cr–ODS_F/M, 12Cr–ODS_F) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained. Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS_F/M steel whereas no distinct temperature dependence for 12Cr–ODS_F steel. Also, there was no significant change in tensile strengths after irradiation below 923 K, but those above 1023 K up to 6.6 × 10 26 n/m 2 ( E > 0.1 MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test.

Friction stir welding was applied to the wrapper tube materials, 11Cr-ferritic/martensitic steel, designed for fast reactors and defect-free welds were successfully produced. The mechanical and microstructural properties of the friction stir welded steel were subsequently investigated. The hardness values of the stir zone were approximately 550 Hv (5.4 GPa) with minimal dependence on the rotational speed, even though they were much higher than those of the base material. However, tensile strengths and elongations of the stir zones were high at 298 K, compared to those of the base material. The excellent tensile properties are attributable to the fine grain formation during friction stir welding.

Out-of-pile high-temperature reaction tests are carried out on the diffusion couples between U–Zr fuel and ODS ferritic steels, in which the concentrations of Cr and Al were systematically changed for improvement of corrosion resistance. The results show that the Al addition to ODS ferritic steel considerably improves the compatibility between U–Zr fuel and the ODS steel. The threshold temperature for reaction layer formation is roughly 50 K higher in the Al-containing ODS ferritic steels than in the steels without Al addition for the testing time up to 900 min. The compatibility improvement mechanism by Al addition is discussed from the viewpoint of activity change. The activity calculation results obtained using general thermodynamic data indicate the possibility that stabilization of the intact α-Zr layer by Al addition is the main mechanism and occupation of steel lattice sites by Al that does not easily diffuse into γ-U–Zr also partly contributes to suppression of the inter-diffusion.

The 9CrODS steel specimens were prepared by different processing with hot-rolling and different cooling rate. The hardness and high-temperature tensile properties were measured. Microstructure was analyzed by means of EBSD inverse pole figure and kernel average miss-orientation angles. The hot-rolled and then air-cooled specimen has the highest tensile strength. The furnace-cooled specimen also has better tensile strength at 700 °C than air-cooled specimen at normalized condition. The high-temperature strength of 9CrODS steel is significantly improved with increasing grain size that can be induced by hot-rolling or furnace-slow cooling, where the localized grain boundary deformation can be suppressed.

This study carried out mechanical tests and microstructural characterizations of several 9Cr and 11Cr-ODS tempered martensitic steels. From those results, the appropriate chemical composition range of 11Cr-ODS tempered martensitic steel was discussed from the viewpoint of high temperature strength improvement. It was shown that the residual α-ferrite fraction in 11Cr-ODS steel was successfully controlled to the same level as the 9Cr-ODS steel, which has excellent high temperature strength, by selecting the chemical compositions on the basis of the multi-component phase diagram. The tensile strength decreased with decreasing W content from 2.0 to 1.4 wt%. On the other hand, creep strength at 973 K did not degrade by the decreasing W content. Both tensile strength and creep strength increased with increasing population of the nano-sized oxide particles. Small angle X-ray scattering analysis revealed that titanium and excess oxygen contents were key parameters in order to improve the dispersion conditions of nano-sized oxide particles.

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower t...

In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burn-up of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed and indicated good application prospects for high burn-up fuel. On the other hand, anomalous irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the anomalous irradiation behaviors.

The recrystallization behavior of 12Cr and 15Cr oxide dispersion-strengthened (ODS) ferritic steels, which are the promising candidate materials for long-life core materials of the advanced fast breeder reactors, was investigated in terms of an intermediate softening heat treatment. It was clarified that keeping recovery structure at the intermediate heat treatment is indispensable for producing recrystallized structure at the final heat treatment. Prevention of repeating recrystallization is owing to the stable {100} 〈110〉 texture formation with less stored strain energy by the cold-rolling of the recrystallized structure. The two-step softening process was proposed to suppress the recrystallization and obtain adequate hardness reduction at the intermediate heat treatment. This process is effective for producing a stable recrystallized structure at the final heat treatment of the manufacturing process of ODS ferritic steel cladding.

The effects of fast neutron irradiation on ring tensile properties of oxide dispersion strengthened (ODS) steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the material irradiation rig at temperatures between 693 and 1108 K to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures. The experimental results showed that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 33 dpa were decreased by about 20%. On the other hand, uniform elongation after irradiation was more than 2% at all irradiation conditions. The ring tensile properties of these ODS claddings remained excellent within these irradiation conditions compared with conventional 11Cr ferritic/martensitic steel (PNC-FMS) claddings.

Hot-rolling at the austenitic γ phase (1000 °C) led to a formation of ultra-fine grains and a high dislocation density within grain interiors by an Ausform process. Those heavily accumulated strains accelerated coarsening of prior austenite grains, and block grains transformed from the coarsened prior austenite grains were also coarser in the subsequent normalizing heat treatment. This process is called as modified Ausform process. The modified Ausform 9CrODS steels were improved in the ultimate tensile strength at 700 °C due to the formation of uniformly coarser block grains.

Tungsten is an effective element for solid-solution strengthening and is a ferrite former in 9CrODS ferritic steel, however excessive tungsten leads to Laves phase formation. Appropriate tungsten concentration in 9CrODS ferritic steel was found to be around 2 mass%, which produces high temperature strength without Laves phase precipitation. The stress increment due to solid-solution strengthening by tungsten was characterized as relatively small, compared to enhanced ferrite formation, the so called residual α-ferrite.

Dual-phase 9Cr-ODS (oxide dispersion-strengthened) steel consisting of residual-α ferrite and α′ martensite has excellent high-temperature strength. This study describes the microstructure of dual-phase 9Cr-ODS steels characterized by atom-probe tomography in order to compare oxide-particle dispersion states in each phase. This revealed that nano-size oxide particles were of the same chemical composition and that their mean size was about 3 nm in each phase. On the other hand, the number density in the residual-α phase was about four times higher than that of the α′ phase. These results indicate that the dense distribution of the oxide particles in the residual-α phase contribute to the excellent high-temperature strength of 9Cr-ODS steel.

Excellent high-temperature strength and resistance to radiation damage of 9Cr Oxide Dispersion Strengthened (9Cr-ODS) martensitic steel have been realized by nano-size Y–Ti–O complex oxide particles dispersed in the matrix and a dual phase structure consisting of α′-martensite and δ-ferrite. These are produced by mechanically alloying Fe–Cr–Ti powders with Y 2 O 3 followed by a hot-consolidation process. Therefore, the hot-consolidation process is the issue to be clarified for the formation of nano-size oxide particle and δ-ferrite. The temperature dependence of the formation and development of nano-size oxide particles and δ-ferrite using mechanically alloyed 9Cr-ODS raw powder were investigated applying X-ray Diffraction and Small Angle X-ray Scattering measurement at SPring-8 and by Electron Probe Micro Analysis. In situ heating measurement techniques with XRD and SAXS enabled real-time observation of phase transformations and allowed correlation between formation of nano-size oxide particle and δ-ferrite.

The process of residual ferrite formation and resultant high-temperature strengthening in 9Cr-ODS ferritic steel was investigated by TEM observation, dilatometric measurement and thermodynamic analysis. Formation of the residual ferrite is dominated by a balance between pinning of α–γ interfaces and the α–γ reverse transformation, and α–γ reverse transformation is affected by dissolution of carbides into the γ-matrix at the AC1 and AC3 points. The fine size of oxide particles is responsible for the higher strength of the residual ferrite containing ODS steels.

The in-reactor creep rupture tests of 11Cr–0.5Mo–2W, V, Nb F/M steel were carried out in the temperature range from 823 to 943 K using materials open test assembly in the fast flux test facility and tensile and temperature-transient-to-burst specimens were irradiated in the experimental fast reactor JOYO at temperatures between 693 and 1013 K to fast neutron doses ranging from 11 to 102 dpa. The results of post-irradiation mechanical tests showed that there was no significant degradation in tensile and transient burst strengths even after neutron irradiation below 873 K, but that there was significant degradation in both strengths at neutron irradiation above 903 K. On the other hand, the in-reactor creep rupture times were equal or greater than those of out-reactor creep even after neutron irradiation at all temperatures. This creep rupture behavior was different from that of tensile and transient burst specimens.

This paper discusses the effects of small amount of Al contamination (

9Cr-ODS (oxide dispersion strengthened) steels developed by JAEA (Japan Atomic Energy Agency) have superior creep properties compared with conventional heat resistant steels. The ODS steels can enormously contribute to practical applications of fast breeder reactors and more attractive fusion reactors. Key issues are developments of material processing procedures for mass production and creep life prediction methods in present R&D. In this study, formulation of creep constitutive equation was performed against the backdrop. The 9Cr-ODS steel displaying an excellent creep property is a dual phase steel. The ODS steel is strengthened by the δ ferrite which has a finer dispersion of oxide particles and shows a higher hardness than the α′ martensite. The δ ferrite functions as a reinforcement in the dual phase 9Cr-ODS steel. Its creep behavior is very unique and cannot be interpreted by conventional theories of heat resistant steels. Alternative qualitative model of creep mechanism was formulated at the start of this study using the results of microstructural observations. Based on the alternative creep mechanism model, a novel creep constitutive equation was formulated using the exponential type creep equation extended by a law of mixture.

Strain-controlled low cycle fatigue tests were conduced for 9Cr–ODS and 12Cr–ODS steels at 873, 923, 973 and 1023 K. The total strain ranges were controlled from 0.5% to 1.5% with strain rate of 0.1%/s. Corresponding plastic strain ranged from 0.01% to 1%. The ODS steels exhibit relatively low level of plastic strain and thus have longer fatigue life in the low total strain region compared to conventional ferritic steels such as Mod. 9Cr–1Mo steel. Neither noticeable cyclic hardening nor softening was observed in ODS steels, whilst Mod. 9Cr–1Mo steel shows an apparent cyclic softening at 873 and 923 K. This is attributable to the excellent microstructure stability due to the presence of nano-sized oxide particles dispersed in ODS steels.

The mechanisms of irradiation embrittlement of two Japanese RAFSs were different from each other. The larger DBTT shift observed in F82H is interpreted by means of both hardening effects and a reduction of cleavage fracture stress by M23C6 carbides precipitation along lath block and packet boundaries, while that of JLF-1 is due to only the hardening effect. Dimensional change measurement during in-pile creep tests revealed the creep strain of F82H was limited at 300 °C. Performance of the weld bond under neutron irradiation will be critical to determine the life time of blanket structural components. Application of the ODS steels, which are resistant to corrosion in supercritical pressurized water, to the water-cooled blanket is essential to increase thermal efficiency of the blanket systems beyond DEMO. The coupling of RAFS and ODS steel could be effective to realize a highly efficient fusion blanket.

The superior creep properties of 9Cr ODS (Oxide Dispersion Strengthened) steels result from small oxide particles dispersed in the matrix. Thus many investigations have studied the creep properties by focusing on the oxide particle distribution, but these results are not sufficient to clarify the creep mechanism and to develop material processing procedures for mass production. Potential microstructural features affecting the creep properties should be studied. The 9Cr ODS steel has many potential microstructural features, for example, PPB (Prior Particle Boundary) pores, precipitates, and two matrix phases. In particular, PPB pores link up and develop into large creep cavities. Therefore, particle size classification of mechanically alloying (MA) produced powders was performed to control PPB. Contrary to our expectations, the particle size classification did not only affect PPB creep cavities, but also other microstructural features. MA powder particle size effects on microstructural features and the creep properties were studied.