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

https://doi.org/10.1016/j.jnucmat.2019.01.052

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.

Measurement of displacement cross sections of aluminum and copper at 5 K by using 200 MeV protons

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

https://doi.org/10.1016/j.jnucmat.2018.05.038

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.

Cavitation damage in double-walled mercury target vessel

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

https://doi.org/10.1016/j.jnucmat.2017.10.044

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.

Model calculation of Cr dissolution behavior of ODS ferritic steel in high-temperature flowing sodium environment

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji

Journal of Nuclear Materials, 505, p.44 - 53, 2018/07

https://doi.org/10.1016/j.jnucmat.2018.03.054

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.

Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

Journal of Nuclear Materials, 499, p.528 - 538, 2018/02

https://doi.org/10.1016/j.jnucmat.2017.11.049

Reaction of hydrogen peroxide with uranium zirconium oxide solid solution; Zirconium hinders oxidative uranium dissolution

Kumagai, Yuta; Takano, Masahide; Watanabe, Masayuki

Journal of Nuclear Materials, 497, p.54 - 59, 2017/12

https://doi.org/10.1016/j.jnucmat.2017.10.050

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.

Effect of chloride ion on corrosion behavior of SUS316L-grade stainless steel in nitric acid solutions containing seawater components under -ray irradiation

Sano, Yuichi; Ambai, Hiromu; Takeuchi, Masayuki; Iijima, Shizuka; Uchida, Naoki

Journal of Nuclear Materials, 493, p.200 - 206, 2017/09

https://doi.org/10.1016/j.jnucmat.2017.06.017

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.

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu; Nakashima, Kazuo*; Kanazawa, Toru*; Tojo, Masayuki*

Journal of Nuclear Materials, 488, p.22 - 32, 2017/05

AA2016-0383.pdf:0.86MB

https://doi.org/10.1016/j.jnucmat.2017.03.004

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.

Thermodynamic evaluation of the solidification phase of molten core-concrete under estimated Fukushima Daiichi Nuclear Power Plant accident conditions

Kitagaki, Toru; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro

Journal of Nuclear Materials, 486, p.206 - 215, 2017/04

AA2016-0278.pdf:0.74MB

https://doi.org/10.1016/j.jnucmat.2017.01.032

Oxygen potentials, oxygen diffusion coefficients and defect equilibria of nonstoichiometric (U,Pu)O

Kato, Masato; Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Akashi, Masatoshi

Journal of Nuclear Materials, 487, p.424 - 432, 2017/04

https://doi.org/10.1016/j.jnucmat.2017.01.056

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.