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The current status and the progress of R&D activities for a Fusion DEMO reactor in the QST Rokkasho Fusion Institute is reported. In order to advance the JA DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach Activity and ITER related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D, design of a fusion neutron source and development of an accelerator, ITER-TBS, tritium handling technology and information technology infrastructures including a supercomputer system and a remote experimentation system are carried out for a Fusion DEMO reactor.

The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokka...

The change in the positron annihilation lifetime (PAL) of vacancy clusters before and after electrolysis hydrogen charging was determined using PAL measurements in electron-irradiated F82H. The experimental change indicated 8 hydrogen atoms were trapped in vacancy clusters; whereas the theoretical calculation resulted in approximately 14 atoms. As the samples were left at room temperature for 5 min until the start of the PAL measurements, the de-trapping effect of hydrogen atoms was also considered; approximately 13 hydrogen atoms were captured at each vacancy cluster. The PAL decreased after annealing at 148 K, which could not be explained theoretically. Therefore, further experiments and discussions are needed to obtain a precise change in the PAL of vacancy clusters containing hydrogen atoms in F82H.

The change in the positron annihilation lifetime (PAL) of vacancy clusters before and after electrolysis hydrogen charging was determined using PAL measurements in electron-irradiated F82H. The experimental change indicated 8 hydrogen atoms were trapped in vacancy clusters; whereas the theoretical calculation resulted in approximately 14 atoms. As the samples were left at room temperature for 5 min until the start of the PAL measurements, the de-trapping effect of hydrogen atoms was also considered; approximately 13 hydrogen atoms were captured at each vacancy cluster. The PAL decreased after annealing at 148 K, which could not be explained theoretically. Therefore, further experiments and discussions are needed to obtain a precise change in the PAL of vacancy clusters containing hydrogen atoms in F82H.

The change in hydrogen thermal desorption characteristic of reduced activation ferritic/martensitic (RAFM) steel for fusion reactor blanket due to creep was investigated to examine the applicability of thermal desorption analysis (TDA) to remaining-life assessment. The hydrogen charging into creep damaged specimens was conducted by means of cathodic electrolysis. The hydrogen-charged samples were subjected to the TDA for the measurement of hydrogen evolution curve. The experimental results revealed that the evolution curve was changed only by creep depending on the test conditions, and the amount of hydrogen desorbed during the analysis (CH) increased with increasing the creep life consumption. A vacancy-type defect and dislocation, which were introduced by the electron beam irradiation and the cold rolling respectively, had a large influence on the desorption characteristic, but they were by no means causes of the change in desorption characteristic with creep. It was also found that the CH was successfully arranged with the parameter derived based on the creep void growth’s law, irrespective of test conditions. The results obtained in this study indicated that the TDA could be a strong tool for the creep damage detection and remaining-life assessment of RAFM steel.

Weibull stress criterion based on the “Local Approach” is applied for rational brittle fracture assessment of steel components taking account of the plastic constraint effect on fracture resistance. Once the cumulative distribution of critical Weibull stress is identified for the material itself, that is Weibull shape parameter m as a material constant independent of size/shape of components, loading mode and temperature, fracture toughness can be corrected to the fracture resistance of structural components on the basis of the Weibull stress criterion. On the other hand, a miniature fracture toughness specimen is useful for evaluating the toughness of local area of the component and limited material such as irradiated test sample. Then, this study focuses on developing a method for identification of shape parameter m of critical Weibull stress distribution using miniature specimens. Two types of miniature 3PB (3-point bend) specimen with thickness B=1.65mm that has different crack depth are used to identify the m-value. The m-value determined from the two types of miniature specimen is found to be almost the same as that obtained from normal size specimens with thickness B=15mm, whereas both testing are conducted at different temperature. It is demonstrated that the fracture toughness of the steel at the other temperature can be precisely corrected to the fracture resistance of the cracked component with lower crack-tip plastic constraint by means of the m-value identified from test results for two types of miniature specimen.

This study assesses the applicability of the Weibull stress for the assessment of the crack-tip plastic constraint effect as well as the mixed mode I and II loading effect on brittle fracture resistance of a ferritic steel. Brittle fracture toughness under different crack-tip plastic constraint conditions and different loading mode conditions was respectively obtained by conducting 3-point bend (3PB) test for single edge cracked specimens with different crack depth subjected to mode I load, and 4-point shear (4PS) test for a single edge cracked specimen with deep crack depth subjected to mixed mode I and II load. By using the results of 3PB tests, the critical Weibull stresses distribution independent of crack-tip plastic constraint was identified, and the critical Weibull stresses obtained by 4PS tests provided significantly smaller distribution than that for 3PB specimen. The mixed mode loading provides the different combined stress field around the crack-tip from that under mode I loading, where the 4PS specimen showed the lower distribution of σ2/σ1 and σ3/σ1 than the 3PB specimens. The lower principal stress ratio to be enlarge the fracture driving force was found to have no or much less influence to linear energy release rate for micro-crack, that was employed in the conventional derivation of the Weibull stress. Consequently, the reason why the conventional Weibull stress could not evaluate the effect of loading mode on fracture resistance could be that the linear energy release rate could not exactly take into account the local fracture driving force enlarged due to the combined stress field under mixed mode loading.

原型炉設計合同特別チームでは、トリチウム増殖ブランケットの概念設計をアクションプランに沿って検討中である。現行設計では筐体内冷却水漏洩事象時に耐圧性を担保できる堅牢性、トリチウム自己生産性、並びに製作性を満足する概念として、円筒型筐体を対象に増殖ブランケットの概念設計検討を進めている。本発表では、円筒型ブランケットにおける概念設計研究の状況と2025年頃までの概念設計フェーズにおける今後の検討課題について報告する。, 第38回プラズマ・核融合学会 年会

Compared to austenitic stainless steels, largely employed in the early fission reactors, high chromium Ferritic/Martenstic (FM) steels, developed since the first half of the 20th century for fossil-fuel power-plants, have a number of advantageous properties among which lower thermal expansion, higher thermal conductivity and better void swelling resistance. At the beginning of the 1970s, FM steels found their first nuclear application as wrapper and fuel cladding materials in sodium-cooled fast reactors. They are now the reference materials for in-vessel components of future fusion reactors, and are considered for in-pile and out-of-pile applications in Generation IV reactors as well as for various other nuclear systems. In this paper, after an introductory historical overview, the challenges associated with the use of FM steels in advanced reactors are addressed, including fabrication, joining and codification issues. The long term evolution of mechanical properties such as the creep and creep-fatigue behaviors is discussed and the degradation phenomena occurring in aggressive environments (lead alloys, high temperature gases, super-critical water and CO2, molten salts) are detailed. The paper also provides a brief overview of the radiation effects in FM steels. The influence of the key radiation parameters e.g. temperature, dose and dose rate on the microstructure and mechanical properties are discussed. The need to better understand the synergistic effects of displacement damage and helium produced by transmutation in fusion conditions is highlighted.

International audience; Compared to austenitic stainless steels, largely employed in the early fission reactors, high chromium Ferritic/Martenstic (FM) steels, developed since the first half of the 20th century for fossil-fuel powerplants, have a number of advantageous properties among which lower thermal expansion, higher thermal conductivity and better void swelling resistance. At the beginning of the 1970s, FM steels found their first nuclear application as wrapper and fuel cladding materials in sodium-cooled fast reactors. They are now the reference materials for in-vessel components of future fusion reactors, and are considered for in-pile and out-of-pile applications in Generation IV reactors as well as for various other nuclear systems. In this paper, after an introductory historical overview, the challenges associated with the use of FM steels in advanced reactors are addressed, including fabrication, joining and codification issues. The long term evolution of mechanical properties such as the creep and creep-fatigue behaviors is discussed and the degradation phenomena occurring in aggressive environments (lead alloys, high temperature gases, super-critical water and CO$_2$, molten salts) aredetailed. The paper also provides a brief overview of the radiation effects in FM steels. The influence of the key radiation parameters e.g. temperature, dose and dose rate on the microstructure and mechanical properties are discussed. The need to better understand the synergistic effects of displacement damage and helium produced by transmutation in fusion conditions is highlighted.

Reduced activation ferritic/martensitic (RAFM) steel, e.g., F82H, is the leading candidate structural material for fusion blanket. Of many blanket concepts, the water-cooled ceramic breeder blanket is an attractive concept because of its compactness and its compatibility with the technologies which using high-temperature water as coolant in conventional light water reactor. For tritium breeding, it is necessary to manage the corrosion of F82H by controlling not chemicals but hydrogen and oxygen in the cooling water. Understanding the effects of oxygen and hydrogen on corrosion property was therefore required for the blanket design. This work aims to investigate the corrosion behavior of F82H in high-temperature water with hydrogen added. Based on the results of the corrosion test, it is speculated that hydrogen addition expected no considerable change of the flow-accelerated corrosion (FAC). In contrast, it was found that the number of cycles to fracture was decreased in the high-temperature water in comparison to the fatigue test at room temperature in air.

SiC/SiC composites are promising structural candidate materials for various nuclear applications over the wide temperature range of 300–1000 °C. Accordingly, irradiation tolerance over this wide temperature range needs to be understood to ensure the performance of these composites. In this study, neutron irradiation effects on dimensional stability and mechanical properties to high doses (11–44 dpa) at intermediate irradiation temperatures (˜600 °C) were evaluated for Hi-Nicalon Type-S or Tyranno-SA3 fiber–reinforced SiC matrix composites produced by chemical vapor infiltration. The influence of various fiber/matrix interfaces, such as a 50–120 nm thick pyrolytic carbon (PyC) monolayer interphase and 70–130 nm thick PyC with a subsequent PyC (˜20 nm)/SiC (˜100 nm) multilayer, was evaluated and compared with the previous results for a thin-layer PyC (˜20 nm)/SiC (˜100 nm) multilayer interphase. Four-point flexural tests were conducted to evaluate post-irradiation strength, and SEM and TEM were used to investigate microstructure. Regardless of the fiber type, monolayer composites showed considerable reduction of flexural properties after irradiation to 11–12 dpa at 450–500 °C; and neither type showed the deterioration identified at the same dose level at higher temperatures (>750 °C) in a previous study. After further irradiation to 44 dpa at 590–640 °C, the degradation was enhanced compared with conventional multilayer composites with a PyC thickness of ˜20 nm. Multilayer composites have shown comparatively good strength retention for irradiation to ˜40 dpa, with moderate mechanical property degradation beginning at 70–100 dpa. Irradiation-induced debonding at the F/M interface was found to be the major cause of deterioration of various composites.

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 ◦C–500 ◦C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

This work addresses an overview of the recent progress, associated risks and development plans of structural materials for in-vessel components (IVCs) in DEMO plants. Reduced Activation Ferritic Martensitic Steels form the primary structural materials for most DEMOs IVCs and will be the focus of the paper. An overview of EU- and J-DEMO programs RAFM steels developments is presented. We highlight the present status in their development for use in DEMO plants, with a focus on the structural materials’ operational, and project-oriented, requirements. The development of materials property handbooks from high-quality data is illustrated. A process to validate these steels for operation in DEMO IVCs is summarised, revealing the pragmatic procedures ongoing, and limitations of this approach due to the synergistic operational effects on IVC materials; the use of in-situ or surveillance monitoring is outlined as a method to accommodate this limitation in synergistic effects and allow confidence in future DEMO operations. The development of advanced modifications to F82H and EUROFER are highlighted, with minor modifications leading to improved low and high-temperature operational design space being open for DEMO reactors.

F82H were irradiated by Kr2+ ions with and without helium up to 1 displacement per atom at 500 °C, and cavity distribution around grain boundary was investigated by transmission electron microscopy. In F82H irradiated by Kr2+ ions, the cavity number density and the cavity size tended to be decreased and increased as a function of the distance from grain boundary, respectively. This gradient of cavity distribution seemed to be depending on the character of grain boundary. While, the size distribution of cavity seemed to be homogeneous in F82H irradiated by Kr2++He+ ions. This indicates that even prior austenite grain boundary hardly acts as sinks for vacancies if a high number density of helium atoms exists, which is probably due to the decrease in vacancy mobility by helium trapping.

低放射化フェライト鋼の高温低サイクル疲労に伴う水素放出特性の変化を調べ，核融合炉ブランケット第一壁の疲労寿命評価への水素昇温脱離分析法の適用性について検討し、その成果について発表を行う。, 日本材料学会九州支部 第７回学術講演会・総会

A water-cooled ceramic breeder blanket system is the Japanese promising design, and the candidate structuralmaterial is a reduced activation ferritic/martensitic steel, F82H. On the other hands, the material of the pipes ofcooling system is planned to an austenitic stainless steel, SUS316 L. The cooling water will commonlyflow bothF82H cooling channels and SUS316 L cooling pipes. Appropriate water chemistry condition for each of them isdifferent, but it is necessary tofind out a reasonable and common water chemistry condition for both of them. Inthe fusion blanket, since the cooling water will be irradiated by neutrons and brought about the radiolysis ofwater, the dissolved oxygen concentration of the cooling water is considered to change by the radiolysis ofwater. This research provides the simultaneous effects of dissolved oxygen and applied stress on surface mor-phology of F82H and SUS316 L by the pressurized water immersion test in dissolved oxygen concentration up to100 ppb.

Reduced activation ferritic/martensitic (RAFM) steel, e.g., F82H, is the leading candidate structural material for fusion blanket. Of many blanket concepts, the water-cooled ceramic breeder blanket is an attractive concept because of its compactness and its compatibility with the technologies which using high-temperature water as coolant in conventional light water reactor. For tritium breeding, it is necessary to manage the corrosion of F82H by controlling not chemicals but hydrogen and oxygen in the cooling water. Understanding the effects of oxygen and hydrogen on corrosion property was therefore required for the blanket design. This work aims to investigate the corrosion behavior of F82H in high-temperature water with hydrogen added. Based on the results of the corrosion test, it is speculated that hydrogen addition expected no considerable change of the flow-accelerated corrosion (FAC). In contrast, it was found that the number of cycles to fracture was decreased in the high-temperature water in comparison to the fatigue test at room temperature in air.

The purpose of this study is to investigate the response of the precipitates against irradiation in F82H steels which are considered as main candidate structure material for the fusion DEMO reactor blanket. M23C6-type precipitates contained F82H-IEA steel, and MX- and M23C6-type precipitates contained F82H-BA12 steel were selected as experimental specimens. Iron irradiation at 673 K using ion accelerator was performed on both steels up to 50 dpa. After irradiation, the instability behavior of precipitates was characterized using transmission electron microscopy. It is found that in F82H-IEA specimens, the number density of radiation defects at 20 dpa was 3.26 × 1021 m−3, and the amorphization of M23C6 was undetected up to 50 dpa. Although Ostwald ripening occurred in M23C6 precipitates, the average size of M23C6 mildly decreased with irradiation dose within this study. For F82H-BA12 specimens, the shrinkage rate of MX- and M23C6-type precipitates was estimated respectively as 0.06 and 0.03 nm/dpa. Hence, the MX and M23C6-type precipitates in F82H steel exhibit an excellent irradiation resistance under the present irradiation condition. Moreover, the behaviors like amorphization and weakening in contrast were not observed in M23C6 precipitates in both specimens, which is different from electron irradiation, presumably due to the factors such as different damage rate. Keywords: F82H, Ion irradiation, Precipitates, Instability

The R&D on SiC/SiC composites under the broader approach (BA) activities between Japan and the EU for fusion DEMO developed a fundamental database of mechanical (Task-1) and physical/chemical (Task-2) properties, with a primary target of the application of SiC/SiC composites as functional structure to be used in the dual coolant breeding blanket concept. This paper aims to summarize previous 10-years activities of the R&D of Japan and to provide the key deliverables toward the DEMO design. In Task-1, good creep and fatigue durability were first demonstrated. Besides, in-plane and inter-laminar strength anisotropy maps at elevated temperatures were comprehensively identified. In parallel, the irradiation effects of SiC materials were specifically determined as input parameters of the analytical model to provide for the irradiation-induced residual stresses. In Task-2, the apparent dose-dependence of the radiation-induced electrical conductivity and the indicative radiation-induced electrical degradation was identified by various irradiation sources. In addition, good gas confinement was identified. Furthermore, no accelerated corrosion for duration of 3000 h at below 1173 K was first demonstrated. With these achievements, it is suggested that the in-vessel component technology, e.g., material corrosion database development, activated corrosion product evaluation code development, compact module tests for validation of the key functions of the components, technology integration assessment for fusion nuclear tests, etc., should be further developed toward DEMO in near-term.

Micro-tensile tests were conducted on reduced-activation ferritic steel F82H specimens after irradiation with Fe and He ions. The displacement damage levels were 50 and 200 dpa, and helium concentrations were 0, 2000 and 5000 appm. The specimens had dimensions of 8 × 1 × 1 µm for the gage section and they were tension tested at room temperature in a vacuum. The yield strengths of the unirradiated specimens were close to the literature value reported for a large, unirradiated specimen. The increases in yield strength and ultimate tensile strength due to Fe ion irradiation were clearly observed. The loss of work hardening was confirmed for the 200 dpa specimens (0 or 2000 appmHe). Higher yield strength and ultimate tensile strength were confirmed for the 50 dpa/5000 appmHe specimen compared to the no helium specimen (50 dpa/0 appmHe). However, no apparent helium effects were confirmed for the 200 dpa/2000 appmHe specimen by the micro-tensile test at room temperature. Keywords: Reduced-activation ferritic steel, Ion irradiation, Helium, Micro-tensile testing

The lead-lithium (Pb-Li) alloy is considered as a coolant and a tritium breeder for fusion reactor blanket systems. One of the critical requirements in realization of the systems is the compatibility of silicon carbide (SiC) and its composites, as structural and/or functional materials of the blanket, with liquid Pb-Li flow at high temperatures. This study aims to clarify the corrosion process of high-purity SiC materials under flow condition of liquid Pb-Li for 3000 h at 700 and 900 °C, respectively, by using the rotating disc corrosion equipment. Furthermore, the effect of wall flow velocity on the compatibility was specifically evaluated. For that purpose the decrease rate (about 32–37%) of the wall flow velocity on disc type’s sample by the dragging effect by fluidic viscosity was first calculated by numerical analysis of the rotating flow condition. The wall flow velocity dependence on the reaction layer’s thickness was then evaluated for both high purity SiC and SiC/SiC composites. The test results show that the formation process of the surface reaction layer mainly depends on both the time required for the chemical reaction and the wall flow velocity in the initial stage but the reaction layer thickness tended to approach to the constant regardless of the wall flow velocity with increasing exposed time. In short, no accelerated corrosion was found for about 3000 h. It was also found that there was no significant difference of the temperatures between 700 and 900 °C. Finally the key corrosion mechanism, i.e., the reaction of oxide impurities, e.g., Li2O, in liquid Pb-Li with the Si-oxide layer over SiC, was identified.

Reduced activation ferritic/martensitic (RAFM) steels are assumed as structural material of blanket design activities. Determining the design limit is essential for the RAFM steels to qualify their use in fusion blankets. This work investigates fatigue properties of a RAFM steel, F82H at elevated temperature up to 823 K in the air and vacuum condition. The tests were carried out with axial strain controlled condition using cylindrical specimen, and the results were summarized using the Manson’s modified universal slope method, which can describe fatigue lifetime as a function of tensile properties. Although fatigue lifetime of F82H decreased with temperature, the fatigue lifetime up to 723 K fell into factor of 2 of the lifetime at ambient temperature. The lifetime at vacuum condition was successfully described using universal slope method. On the other hand, the lifetime in the air was 1/3 or shorter than that in vacuum condition. It is notable that a heat of F82H prepared for IEA round robin tests demonstrated 1/5 of shorter lifetime than latest F82H. The difference is discussed with inclusion morphologies affected by steel manufacturing process and so on. This report has been prepared as an account of work assigned to the Japanese Implementing Agency within the ‘‘Broader Approach Agreement’’ between the Government of Japan and the European Atomic Energy Community.

The requirements for Japanese DEMO are to demonstrate (1) steady and stable electrical power generation, (2) plant availability on a level relevant for commercialization, and (3) self-sufficiency of fuels. The DEMO design activity has been mainly focused on challenging DEMO design issues such as divertor heat removal, superconducting magnets, breeding blankets, remote maintenance and the water cooling system, in order to reach the systemic consistency toward a promising DEMO concept. The current status of the DEMO conceptual design activity is presented.

Room temperature micro-tensile tests were successfully performed on F82H specimens neutron-irradiated at 573 K up to 5 dpa and unirradiated using a focused ion beam (FIB) device. Dimensions of the gauge section for micro-tensile specimens were about 10 µm (length) × 1 µm2 (area). These specimens were fabricated in a FIB device. The tensile properties obtained on specimens of micrometer size qualitatively agreed with results from millimeter size specimens. Using this micro-tensile testing technique is considered to become very useful data in combining the micro-tensile data on ion irradiated F82H with data on neutron irradiated material. This method overcomes the limitation on obtaining mechanical property on ion irradiated specimens to only micro-indentation testing. Keywords: Ultra-small testing technology, Micro-tensile testing, Neutron irradiation, Reduced activation ferritic/martensitic steel

Preparation of irradiation database is essential for the design activity of fusion reactor in-vessel components since it is difficult to suppress the irradiation-induced degradation of material properties within the negligible level. On the other hand, the preparation of irradiation database corresponding to the real fusion in-vessel environment in the level that the existing design codes require for new material is not realistic. This limitation is due to the limited capability of irradiation experiments, and the fusion neutron irradiation environment is not available for the next ten years. The probability-based design method could be a potential alternative approach to evaluating the possibility of components’ failure, where the material property is defined with a probability density function. The statistical analyses of data are required to provide the function, and a case study on as prepared and post-irradiation tensile properties of Japanese reduced activation ferritic/martensitic steel F82H was conducted. It was indicated that the most data show a normal distribution, but post-irradiation elongation data shows a possibility to follow a Weibull distribution.