Negyesi, M.; Amaya, Masaki
Journal of Nuclear Materials, 524, p.263 - 277, 2019/10
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
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 -phase to the -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.
Nordlund, K.*; Zinkle, S. J.*; Sand, A. E.*; Granberg, F.*; Averback, R. S.*; Stoller, R. E.*; Suzudo, Tomoaki; Malerba, L.*; Banhart, F.*; Weber, W. J.*; et al.
Journal of Nuclear Materials, 512, p.450 - 479, 2018/12
Scientific understanding of any kind of radiation effects starts from the primary damage. We consider the extensive experimental and computer simulation studies that have been performed over the past several decades on what the nature of the primary damage is. We review both the production of crystallographic or topological defects in materials as well as radiation mixing, i.e. the process where atoms in perfect crystallographic positions exchange positions with other ones in non-defective positions. We also consider the recent effort to provide alternatives to the current international standard for quantifying this energetic particle damage, the Norgett-Robinson-Torrens displacements per atom (NRT-dpa) model for metals. We present in detail new complementary displacement production estimators ("athermal recombination corrected dpa": arc-dpa) and atomic mixing ("replacements per atom": rpa) functions that extend the NRT-dpa, and discuss their advantages and limitations.
Iwamoto, Yosuke; Yoshida, Makoto*; Yoshiie, Toshimasa*; Satoh, Daiki; Yashima, Hiroshi*; Matsuda, Hiroki; Meigo, Shinichiro; Shima, Tatsushi*
Journal of Nuclear Materials, 508, p.195 - 202, 2018/09
To validate the displacement damage model in radiation transport codes used for the estimation of radiation damages at accelerator facilities, we measured electrical resistance increase of aluminum and copper induced by radiation defects under the cryogenic 200 MeV proton irradiation. The irradiation device had the structure to cool two irradiation samples at same time using thermal conductance. The aluminum and copper wire with 250 m diameter was sandwiched between two AlN plates with excellent thermal conductivity and electrical insulation. As a result, temperature of irradiation samples was kept at below 5 K under proton irradiation with beam intensity below 3 nA. The experimental displacement cross section agreed with calculated results with defect production efficiency.
Naoe, Takashi; Xiong, Z.*; Futakawa, Masatoshi
Journal of Nuclear Materials, 506, p.12 - 18, 2018/08
A mercury target for neutron source (made of 316L SS) suffers not only proton and neutron radiation damage, but also cyclic impact stress caused by pressure waves. In the previous study, we carried out an ultrasonic fatigue test to investigate the gigacycle fatigue strength of 316L SS, concluding that specimen surface temperature rose abruptly more than 300C just before failure. In this study, to clarify the mechanism of the temperature rise, we measured temperature distribution with a thermography during the fatigue test. The experimental results showed that the temperature rose locally only at the crack tip and the peak position moved with the crack propagation. We also carried out a nonlinear structural analysis by LS-DYNA to estimate the temperature rise with strain energy of elements. The analytical result showed that the heat due to plastic deformation at the crack tip is dominant for the temperature rise rather than the friction between crack surface.
Naoe, Takashi; Wakui, Takashi; Kinoshita, Hidetaka; Kogawa, Hiroyuki; Haga, Katsuhiro; Harada, Masahide; Takada, Hiroshi; Futakawa, Masatoshi
Journal of Nuclear Materials, 506, p.35 - 42, 2018/08
A mercury target vessel made of 316L SS is damaged due to the cavitation caused by the pressure waves in mercury. Cavitation damage reduces the structural integrity of the target front, called "beam window", being major factor to determine the lifetime of target vessel. Aiming at mitigating the cavitation damage by faster mercury flow in narrow channel, we employed a target vessel with a double-walled structure at the beam window along with a gas microbubbles injection. After operating the double-walled target vessel with a beam power of 300 to 500 kW, we cut out the beam window using an annular cutter to examine the damage inside it, and found that damages with maximum pit depth of approximately 25 m distributed in a belt on the specimen facing narrow channel. Furthermore, numerical simulation result showed that the distribution of negative pressure period from beam injection to 1 ms was correlated with the damage distribution in the narrow channel. It was suggested that the cavitation induced by relatively short negative pressure period contributed to the damage formation.
Suzudo, Tomoaki; Tsuru, Tomohito; Hasegawa, Akira*
Journal of Nuclear Materials, 505, p.15 - 21, 2018/07
Tungsten (W) is considered as a promising candidate for plasma-facing materials for future nuclear fusion devices, and selecting optimal alloying constituents is a critical issue to improve radiation resistance of the W alloys as well as to improve their mechanical properties. We conducted in the current study a series of first-principles calculations for investigating solvent-solute mixed dumbbells in W crystals. The results suggested that titanium (Ti), vanadium (V), and chromium (Cr) are favorable as solutes for W alloys from irradiation-effect perspectives because these elements are expected to promote vacancy-interstitial recombination without causing radiation-induced precipitation that reduces ductility of irradiated materials.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji
Journal of Nuclear Materials, 505, p.44 - 53, 2018/07
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.
Muta, Hiroaki*; Nishikane, Ryoji*; Ando, Yusuke*; Matsunaga, Junji*; Sakamoto, Kan*; Harjo, S.; Kawasaki, Takuro; Oishi, Yuji*; Kurosaki, Ken*; Yamanaka, Shinsuke*
Journal of Nuclear Materials, 500, p.145 - 152, 2018/03
Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki
Journal of Nuclear Materials, 499, p.528 - 538, 2018/02
Kobata, Masaaki; Okane, Tetsuo; Nakajima, Kunihisa; Suzuki, Eriko; Owada, Kenji; Kobayashi, Keisuke*; Yamagami, Hiroshi; Osaka, Masahiko
Journal of Nuclear Materials, 498, p.387 - 394, 2018/01
In this study, for the understandings of Cesium (Cs) adsorption behavior on structure materials in severe accidents at a light water nuclear reactor, the chemical state of Cs and its distribution on the surface of SUS304 stainless steel (SS) with different Si concentration were investigated by hard X-ray photoelectron spectroscopy (HAXPES) and scanning electron microscope / energy dispersive X-ray spectroscopy (SEM/EDX). As a result, it was found that Cs is selectively adsorbed at the site where Si distributes with high concentration. CsFeSiO is a dominant Cs products in the case of low Si content, mainly formed, while CsSiO and CsSiO are formed in addition to CsFeSiO in the case of high Si content. The chemical forms of the Cs compounds produced in the adsorption process on the SS surface has a close correlation with the concentration and chemical states of Si originally included in SS.
Kumagai, Yuta; Takano, Masahide; Watanabe, Masayuki
Journal of Nuclear Materials, 497, p.54 - 59, 2017/12
We studied oxidative dissolution of uranium and zirconium oxide [(U,Zr)O] in aqueous HO solution. The interfacial reaction is essential for anticipating how a (U,Zr)O-based molten fuel may chemically degrade after a severe accident under influence of ionizing radiation. We conducted our experiments with (U,Zr)O powder and quantitated the HO reaction via dissolved U and HO concentrations. The dissolution yield relative to HO consumption was far less for (U,Zr)O compared to that of UO. The reaction kinetics indicates that most of the HO catalytically decomposed to O at the surface of (U,Zr)O. We confirmed the HO catalytic decomposition via O production (quantitative stoichiometric agreement). In addition, post-reaction Raman scattering spectra of the undissolved (U,Zr)O showed no additional peaks (indicating a lack of secondary phase formation). The (U,Zr)O matrix is much more stable than UO against HO-induced oxidative dissolution.
Tanno, Takashi; Takeuchi, Masayuki; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Materials, 494, p.219 - 226, 2017/10
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 95C. The corrosion rate exponentially decreased with effective solute chromium concentration (Cr) 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 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, concentrated nitric acid and addition of oxidizing ions helped to prevent the active dissolution, and suppressed the corrosion rate.
Sano, Yuichi; Ambai, Hiromu; Takeuchi, Masayuki; Iijima, Shizuka; Uchida, Naoki
Journal of Nuclear Materials, 493, p.200 - 206, 2017/09
Concerning the Fukushima Daiichi Nuclear Power Plant accident, we investigated the effect of chloride ion on the corrosion behavior of SUS316L stainless steel, which is a typical material for the equipment used in reprocessing, in HNO solution containing seawater components, including under the -ray irradiation condition. Electrochemical and immersion tests were carried out using a mixture of HNO and artificial seawater (ASW). In the HNO solution containing high amounts of ASW, the cathodic current densities increased and uniform corrosion progressed. This might be caused by strong oxidants, such as Cl and NOCl, generated in the reaction between HNO and Cl ions. The corrosion rate decreased with the immersion time at low concentrations of HNO, while it increased at high concentrations. Under the -ray irradiation condition, the corrosion rate decreased due to the suppression of the cathodic reactions by the reaction between the above oxidants and HNO generated by radiolysis.
Nakajima, Kunihisa; Takai, Toshihide; Furukawa, Tomohiro; Osaka, Masahiko
Journal of Nuclear Materials, 491, p.183 - 189, 2017/08
One of the main chemical forms of cesium in the gas phase during severe accidents of light water reactor is expected to be cesium metaborate, CsBO, by thermodynamic equilibrium calculation considering reaction with boron. But accuracy of the thermodynamic data of gaseous metaborate, CsBO(g), has been judged as poor quality. Thus, Knudsen effusion mass spectrometric measurement of CsBO was carried out to obtain reliable thermodynamic data. The evaluated values of standard enthalpy of formation of CsBO(g), H(CsBO,g), by the 2nd and 3rd law treatments are -700.710.7 kJ/mol and -697.010.6 kJ/mol, respectively, and agree with each other within the errors, which suggests our data are reliable. Further, it was found that the existing data of the Gibbs energy function and the standard enthalpy of formation agreed well with the values evaluated in this study, which indicates the existing thermodynamic data are also reliable.
Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu; Nakashima, Kazuo*; Kanazawa, Toru*; Tojo, Masayuki*
Journal of Nuclear Materials, 488, p.22 - 32, 2017/05
The authors previously conducted thermogravimetric analyses on zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study.
Miwa, Shuhei; Osaka, Masahiko
Journal of Nuclear Materials, 487, p.1 - 4, 2017/04
Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO specimen were determined. The oxidation and reduction rates of the MgO-PuO were found to be low compared with those of PuO. It is note that the changes in O/Pu ratios of MgO-PuO from stoichiometry were smaller than those of PuO at high oxygen partial pressure. From these results, it can be said that MgO matrix lower the oxygen supply and release of PuO, which is preferable as the minor actinides incineration devices, since the high oxygen potentials of minor actinide oxides can cause certain problems in terms of thermochemical aspects such as enlarged cladding inner-surface corrosion.
Kitagaki, Toru; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro
Journal of Nuclear Materials, 486, p.206 - 215, 2017/04
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.
Journal of Nuclear Materials, 487, p.229 - 237, 2017/04
Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200C. This degradation was attributed to grain boundary sliding deformation with / transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 C unlike the other tested materials. Present study includes the result of "R&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
Kato, Masato; Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Akashi, Masatoshi
Journal of Nuclear Materials, 487, p.424 - 432, 2017/04
Oxygen potential of (U,Pu)O was evaluated based on defect chemistry using an updated experimental data set. The relationship between oxygen partial pressure and deviation in (U,Pu)O was analyzed, and equilibrium constants of defect formation were determined as functions of Pu content and temperature. Brouwer's diagrams were constructed using the determined equilibrium constants, and a relational equation to determine O/M ratio was derived as functions of O/M ratio, Pu content and temperature. In addition, relationship between oxygen potential and oxygen diffusion coefficients were described.