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Journal Articles

Tensile properties of irradiated modified 316 stainless steel (PNC316) at slow strain rates

Yano, Yasuhide; Miyazawa, Takeshi; Tanno, Takashi; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Kaito, Takeji; Otsuka, Satoshi

Journal of Nuclear Science and Technology, 8 Pages, 2025/00

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

The effects of strain rate on tensile properties of irradiated modified 316 stainless steel (PNC316) claddings were investigated. PNC316 claddings were irradiated at the experimental fast reactor Joyo using CRT402 control rod assembly at 400$$^{circ}$$C up to 25 dpa. Post-irradiation ring tensile tests were carried out at strain rates of 3.3$$times$$10$$^{-6}$$, 3.3$$times$$10$$^{-7}$$ and 3.3$$times$$10$$^{-8}$$ s$$^{-1}$$ at a test temperature of 350$$^{circ}$$C. The results showed no obvious dependence of all strain rates on tensile properties, although a slight decrease in total elongation was observed at the slowest strain rate of 3.3$$times$$10$$^{-8}$$ s$$^{-1}$$. In addition, only a part of fracture surface at the slowest strain rate showed intergranular type region in the inner surface area, although the grain boundary separation occurred on inner surfaces near the fracture region at all strain rates. It is suggested that presence of a high content of helium near the inner surfaces would be related to the fracture behavior.

JAEA Reports

High-temperature strength of modified type 316 steel for fast reactor fuel before and after neutron irradiation

Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji

JAEA-Technology 2024-009, 140 Pages, 2024/10

JAEA-Technology-2024-009.pdf:8.03MB

For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900$$^{circ}$$C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.

Journal Articles

Oxide particles in oxide dispersion strengthened steel neutron-irradiated up to 158 dpa at Joyo

Toyama, Takeshi*; Tanno, Takashi; Yano, Yasuhide; Inoue, Koji*; Nagai, Yasuyoshi*; Otsuka, Satoshi; Miyazawa, Takeshi; Mitsuhara, Masatoshi*; Nakashima, Hideharu*; Onuma, Masato*; et al.

Journal of Nuclear Materials, 599, p.155252_1 - 155252_14, 2024/10

 Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)

We investigated the stability of oxide nano particles in oxide dispersion-strengthened (ODS) steel, which is a promising candidate material for next-generation reactors, under neutron irradiation at high temperature to high doses. MA957, a 14Cr-ODS steel, was irradiated with Joyo in Japan Atomic Energy Agency under irradiation conditions of 130 dpa at 502$$^{circ}$$C, 154 dpa at 589$$^{circ}$$C, and 158 dpa at 709$$^{circ}$$C. Three-dimensional atom probe (3D-AP) and transmission electron microscope (TEM) observation were performed to characterize the oxide particles in the ODS steels. A high number density of Y-Ti-O particle was observed in the unirradiated and irradiated samples. Almost no change in the morphology of the oxide particles, i.e. average diameter, number density, and chemical composition, has been observed in the samples irradiated to 130 dpa at 502$$^{circ}$$C and to 154 dpa at 589$$^{circ}$$C. A slight decrease in number density was observed in the sample irradiated to 158 dpa at 709$$^{circ}$$CC. The hardness of any of the irradiated samples was almost unchanged from that of the unirradiated sample. It was revealed that the oxide particles existed stable, and the strength of the material was sufficiently maintained even after being neutron irradiated to high dose of $$sim$$160 dpa at high temperature up to 700$$^{circ}$$C. A part of this study includes the results of MEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0219214482.

Journal Articles

Chapter 9, Advanced materials; Oxide-dispersion strengthened steels

Otsuka, Satoshi; Tanno, Takashi; Yano, Yasuhide; Kaito, Takeji

Materials and Processes for Nuclear Energy Today and in the Future, p.279 - 297, 2024/10

The oxide dispersion strengthening is an effective technique for improving the mechanical strength of the steel. The dispersed oxides prevent the gliding motion of dislocations, thus remarkably enhancing the resistance to high-temperature deformation and rupture of steels. Extensive efforts have been made to develop ODS steels in the fields of nuclear and fusion engineering. Research has been done to improve their performance and meet the requirements such as irradiation resistance, high-temperature strength, and corrosion resistance. Based on recent research, the high-density dispersion of nanosized oxides could improve the irradiation resistance of the steels in addition to high-temperature strength because the interface between oxide and matrix could act as sink sites for point defects. This section overviews the ODS steel development for nuclear application.

Journal Articles

Formulation of high-temperature strength equation of 9Cr-ODS tempered martensitic steels using the Larson-Miller parameter and life-fraction rule for rupture life assessment in steady-state, transient, and accident conditions of fast reactor fuel

Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Toyama, Takeshi*; Onuma, Masato*; et al.

Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05

 Times Cited Count:1 Percentile:57.00(Materials Science, Multidisciplinary)

Journal Articles

Tensile properties of modified 316 stainless steel (PNC316) after neutron irradiation over 100 dpa

Yano, Yasuhide; Uwaba, Tomoyuki; Tanno, Takashi; Yoshitake, Tsunemitsu; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Science and Technology, 61(4), p.521 - 529, 2024/04

 Times Cited Count:3 Percentile:62.75(Nuclear Science & Technology)

The effects of fast neutron irradiation on tensile properties of modified 316 stainless steel (PNC316) claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400 and 735 $$^{circ}$$C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400 and 500$$^{circ}$$C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.

Journal Articles

High-temperature creep properties of 9Cr-ODS tempered martensitic steel and quantitative correlation with its nanometer-scale structure

Otsuka, Satoshi; Shizukawa, Yuta; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Onizawa, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.

Journal of Nuclear Science and Technology, 60(3), p.288 - 298, 2023/03

 Times Cited Count:4 Percentile:53.26(Nuclear Science & Technology)

JAEA has been developing 9Cr-oxide dispersion strengthened (ODS) tempered martensitic steel(TMS) as a candidate material for the fuel cladding tubes of sodium-cooled fast reactors(SFRs). The reliable prediction of in-reactor creep-rupture strength is critical for implementing the 9Cr-ODS TMS cladding tube in the SFR. This study investigated the quantitative correlation between the creep properties of 9Cr-ODS TMS at 700 $$^{circ}$$C and the dispersions of nanosized oxides by analyzing the creep data and the material's nanostructure. The possibility of deriving a formula for estimating the in-reactor creep properties of 9Cr-ODS TMSs based on an analysis of the nanostructure of neutron-irradiated 9Cr-ODS TMSs was also discussed. The creep properties of 9Cr-ODS TMS at 700 $$^{circ}$$C closely correlated with the dispersion of nanosized oxide particles. The correlation between creep-rupture lives and nanosized oxide particle dispersion was determined using existing creep models. The elucidation of correlation between the stress exponent of secondary creep rate and the nanostructure is essential to enhance future modeling reliability and formulation.

Journal Articles

High temperature mechanical properties and microstructure in 9Cr or 12Cr oxide dispersion strengthened steels

Mitsuhara, Masatoshi*; Kurino, Koichi*; Yano, Yasuhide; Otsuka, Satoshi; Toyama, Takeshi*; Onuma, Masato*; Nakashima, Hideharu*

Tetsu To Hagane, 109(3), p.189 - 200, 2023/03

 Times Cited Count:1 Percentile:12.80(Metallurgy & Metallurgical Engineering)

Oxide Dispersion Strengthened (ODS) ferritic steel, a candidate material for fast reactor fuel cladding, has low thermal expansion, good thermal conductivity, and excellent resistance to irradiation damage and high temperature strength. The origin of the excellent high-temperature strength lies in the dispersion of fine oxides. In this study, creep tests at 700 or 750$$^{circ}$$C, which are close to the operating temperatures of fast reactors, and high-temperature tensile tests at 900 to 1350 $$^{circ}$$C, which simulate accident conditions, were conducted on 9Cr ODS ferritic steels, M11 and MP23, and 12Cr ODS ferritic steel, F14, to confirm the growth behavior of oxides. In the M11 and F14 creep test samples, there was little oxide growth or decrease in number density from the initial state, indicating that dispersion strengthening by oxides was effective during deformation. After creep deformation of F14, the development of dislocation substructures such as dislocation walls and subgrain boundaries was hardly observed, and mobile dislocations were homogeneously distributed in the grains. The dislocation density increased with increasing stress during the creep test. In the high-temperature ring tensile tests of MP23 and F14, the strength of both steels decreased at higher temperatures. In MP23, elongation decreased with increasing test temperature from 900 to 1100 $$^{circ}$$C, but increased at 1200 $$^{circ}$$C, decreased drastically at 1250 $$^{circ}$$C, and increased again at 1300 $$^{circ}$$C. In F14, elongation decreased with increasing temperature. It was inferred that the formation of the $$delta$$-ferrite phase was responsible for this complex change in mechanical properties of MP23 from 1200 to 1300 $$^{circ}$$C.

Journal Articles

Effect of nitrogen concentration on creep strength and microstructure of 9Cr-ODS ferritic/martensitic steel

Oka, Hiroshi*; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Hashimoto, Naoyuki*

Journal of Nuclear Materials, 572, p.154032_1 - 154032_8, 2022/12

 Times Cited Count:5 Percentile:61.74(Materials Science, Multidisciplinary)

9Cr oxide dispersion strengthened steels with slightly different nitrogen concentrations (0.0034 - 0.029 wt%) were prepared and their creep property at 973 K was investigated with microstructural characterization before and after the creep test. The creep strength decreased significantly as the nitrogen concentration increased. Microstructural observation revealed that, in the higher nitrogen concentration specimen, coarse Y-rich inclusions were found along the boundary between transformed ferrite region and residual ferrite region. The solubility difference of nitrogen in $$alpha$$ and $$gamma$$ phase would induce the localized increment of nitrogen concentration in the boundary region during the austenitizing process, resulting in the thermodynamic destabilization and subsequent coarsening of the dispersed oxide particles. The rows of creep voids were found near the rupture part of the crept specimen, suggesting that the coarse inclusions were the starting point of creep void formation and the subsequent premature fracture.

Journal Articles

Development and issues of fast reactor core materials

Kaito, Takeji; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi

Nuclear Materials Letters (Internet), p.29 - 43, 2022/12

no abstracts in English

JAEA Reports

Evaluation of tensile and creep properties on 9Cr-ODS steel claddings

Yano, Yasuhide; Hashidate, Ryuta; Tanno, Takashi; Imagawa, Yuya; Kato, Shoichi; Onizawa, Takashi; Ito, Chikara; Uwaba, Tomoyuki; Otsuka, Satoshi; Kaito, Takeji

JAEA-Data/Code 2021-015, 64 Pages, 2022/01

JAEA-Data-Code-2021-015.pdf:2.6MB

From a view point of practical application of fast breeder reactor cycles, which takes advantage of safety and economic efficiency and makes a contribution of volume reduction and mitigation of degree of harmfulness of high-level radioactive waste, it is necessary to develop fuel cladding materials for fast reactors (FRs) in order to achieve high-burnup. Oxide dispersion strengthened (ODS) steel have been studied for use as potential fuel cladding materials in FRs owing to their excellent resistance to swelling and their high-temperature strength in Japan Atomic Energy Agency. It is very important to establish the materials strength standard in order to apply ODS steels as a fuel cladding. Therefore, it is necessary to acquire the mechanical properties such as tensile, creep rupture strength tests and so on. In this study, tensile and creep rupture strengths of 9Cr-ODS steel claddings were evaluated using by acquired these data. Because of the phase transformation temperature of 9Cr-ODS steel, temperature range for the evaluation was divided into two ones at AC1 transformation temperature of 850$$^{circ}$$C.

Journal Articles

Tensile properties on dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging

Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Materials, 555, p.153105_1 - 153105_8, 2021/11

 Times Cited Count:3 Percentile:10.00(Materials Science, Multidisciplinary)

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$$^{circ}$$C up to 30,000 h. Characterization of microstructure was carried out by scanning electron microscopy and transmission electron microscopy. Microstructural analysis showed that the microstructure in the weld metals consisted of lath martensite containing a small amount of residual austenite. Thermal aging hardening of WMs occurred at 400 and 450$$^{circ}$$C due to the effects of both a-a' 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.

Journal Articles

Laser beam direct energy deposition of graded austenitic-to-martensitic steel junctions compared to dissimilar electron beam welding

Villaret, F.*; Boulnat, X.*; Aubry, P.*; Yano, Yasuhide; Otsuka, Satoshi; Fabregue, D.*; de Carlan, Y.*

Materials Science & Engineering A, 824, p.141794_1 - 141794_10, 2021/09

 Times Cited Count:4 Percentile:22.23(Nanoscience & Nanotechnology)

Journal Articles

Effects of thermal aging on the mechanical properties of FeCrAl-ODS alloy claddings

Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Kaito, Takeji; Ukai, Shigeharu*

Materials Transactions, 62(8), p.1239 - 1246, 2021/08

 Times Cited Count:10 Percentile:46.82(Materials Science, Multidisciplinary)

The FeCrAl-ODS alloy claddings were manufactured and Vickers hardness, ring tensile tests and TEM observations of these claddings were performed to investigate the effects of thermal aging at 450 $$^{circ}$$C for 5,000 and 15,000 h. The age-hardening of all FeCrAl-ODS alloy cladding was found. In addition, the significant increase in tensile strength was accompanied by much larger loss of ductility. It was suggested that this age-hardening behavior was attributed to the (Ti, Al)-enriched phase ($$beta$$' phase) and the $$alpha$$' phase precipitates (content of Al is $$<$$ 7 wt%). In comparison with FeCrAl-ODS alloys with almost same chemical compositions, there was significant age-hardening in both alloys. However, the extrusion bar with no-recrystallized structures was keeping good ductility. It was suggested that this different behavior of reduction ductility was attributed to the effects of grain boundaries, dislocation densities and specimen preparation direction.

Journal Articles

Solid-solution strengthening by Al and Cr in FeCrAl oxide-dispersion-strengthened alloys

Ukai, Shigeharu*; Yano, Yasuhide; Inoue, Toshihiko; Sowa, Takashi*

Materials Science & Engineering A, 812, p.141076_1 - 141076_11, 2021/04

 Times Cited Count:23 Percentile:78.65(Nanoscience & Nanotechnology)

FeCrAl oxide dispersion strengthened alloys are promising materials for accident tolerant fuels for light water reactors (LWRs). In these alloys, Al and Cr are key elements with important synergistic effects: enhancement of the formation of oxidation-resistant Al$$_{2}$$O$$_{3}$$ phase by Cr addition and suppression of the formation of the embrittling Cr-rich $$alpha$$' phase by Al addition. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The Al and Cr contents were systematically varied from 9-16 at.% and 10-17 at.%, respectively, and tensile tests were conducted at 298 K, 573 K and 973 K in the as-annealed condition. The solid solution strengthening increased linearly, 20 MPa per 1 at.% Al and 5 MPa per 1 at.% Cr, at the typical LWR operational temperature of 573 K. The conventional Fleischer-Friedel and Labusch theories cannot explain this level of solid-solution strengthening. It was shown that Suzuki's double kink theory for screw dislocations reasonably predicts the solid solution strengthening by Al and Cr as well as the inverse dependency on the absolute temperature and linear dependency on the Al and Cr content.

Journal Articles

Measurement of niobium reaction rate for material surveillance tests in fast reactors

Ito, Chikara; Maeda, Shigetaka; Inoue, Toshihiko; Tomita, Hideki*; Iguchi, Tetsuo*

Radiation Protection, 40(6), p.491 - 495, 2020/11

A highly accurate and precise technique for measurement of the $$^{93}$$Nb(n,n')$$^{93m}$$Nb reaction rate was established for the material surveillance tests, etc. in fast reactors. The self-absorption effect on the measurement of the characteristic X-rays emitted by $$^{93m}$$Nb was decreased by the solution and evaporation to dryness of niobium dosimeter. A highly precise count of the number of $$^{93}$$Nb atoms was obtained by measuring the niobium solution concentration using inductively coupled plasma mass spectrometry. X-rays of $$^{93m}$$Nb were measured accurately by means of comparing the X-ray intensity of irradiated niobium solution with that of the solution in which stable $$^{93}$$Nb was added. The difference between both intensities indicates the effect of $$^{182}$$Ta, which is generated from an impurity tantalum, and the intensity of X-rays from $$^{93m}$$Nb was evaluated. Measurement error of the $$^{93}$$Nb(n,n')$$^{93m}$$Nb reaction rate was reduced to be less than 4%, which was equivalent to the other reaction rate errors of dosimeters used for Joyo dosimetry. In addition, an advanced technique using Resonance Ionization Mass Spectrometry was proposed for the precise measurement of $$^{93m}$$Nb yield, and $$^{93m}$$Nb will be resonance-ionized selectively by discriminating the hyperfine splitting of the atomic energy levels between $$^{93}$$Nb and $$^{93m}$$Nb at high resolution.

Journal Articles

Oxide dispersion strengthened steels

Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi

Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08

Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.

Journal Articles

Concerning aging of nuclear fuel material use facilities Examination of measures to improve safety assessment methods

Sakamoto, Naoki; Fujishima, Tadatsune; Mizukoshi, Yasutaka

Hozengaku, 19(2), p.125 - 126, 2020/07

The five post-irradiation examination facilities in JAEA's Oarai research and development institute have been operated for over 40 years in order to investigate the irradiation performance of fast reactor fuel materials. The equipment associated with these facilities has been managed to maintain secure from the problems occurred in the process of aging. Therefore, we established a safety assessment method for aging facilities in 2002, and we have been conducting maintenance management of facilities since then. In this study, improvement plans of the safety assessment method are considered in order to solve the issues detected as a result of analysis of past maintenance information.

Journal Articles

Effect of nickel concentration on radiation-induced diffusion of point defects in high-nickel Fe-Cr-Ni model alloys during neutron and electron irradiation

Sekio, Yoshihiro; Sakaguchi, Norihito*

Materials Transactions, 60(5), p.678 - 687, 2019/05

 Times Cited Count:8 Percentile:35.86(Materials Science, Multidisciplinary)

The quantitative evaluation of vacancy migration energies in high nickel model alloy was conducted by analyzing the void denuded zone (VDZ) width formed near grain boundaries under neutron and electron irradiation. The microstructures of Fe-15Cr-xNi (x=15, 20, 25, 30 mass%) alloys that were neutron irradiated at 749 K and electron irradiated at 576 K-824 K were examined. The VDZ widths increased with increasing Ni content in both irradiation experiments, which implies an increase of the vacancy mobility. The vacancy migration energies were estimated from the temperature dependence of the VDZ widths, and the energies were 1.09, 0.97, 0.90, and 0.77 eV for the alloys containing 15, 20, 25, and 30 mass% Ni, respectively. From the obtained energies, the effective vacancy diffusivity and excess vacancy concentration were estimated using the analytical equation of the VDZ width, which quantitatively confirmed the increase of the vacancy mobility with increasing Ni content.

Journal Articles

Ultra-high temperature creep rupture and transient burst strength of ODS steel claddings

Yano, Yasuhide; Sekio, Yoshihiro; Tanno, Takashi; Kato, Shoichi; Inoue, Toshihiko; Oka, Hiroshi; Otsuka, Satoshi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; et al.

Journal of Nuclear Materials, 516, p.347 - 353, 2019/04

 Times Cited Count:21 Percentile:88.33(Materials Science, Multidisciplinary)

9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 $$^{circ}$$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 $$^{circ}$$C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 $$^{circ}$$C. This superior strength seemed to be owing to transformation of the matrix from the $$alpha$$-phase to the $$gamma$$-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.

183 (Records 1-20 displayed on this page)