Large-scale atomistic simulations of cleavage in BCC Fe using machine-learning potential

Suzudo, Tomoaki; Ebihara, Kenichi; Tsuru, Tomohito; Mori, Hideki*

Zairyo, 73(2), p.129 - 135, 2024/02

https://doi.org/10.2472/jsms.73.129

Body-centered-cubic transition metals, such as Fe and W, cleave along the {100} plane. To find out the mechanism of this response, atomistic simulations of curved crack-fronts of bcc Fe were conducted at 0 K using an interatomic potential created by an artificial neural network (ANN) technique. We discovered that dislocations can be emitted from the curved crack fronts along the {110} crack plane, and this phenomenon explains why the cleavage is observed only along the {100} plane. In addition, the cleavage simulations along {100} at the elevated temperature were found to be accompanied by plasticity; namely, they represented more realistic fracture.

Cleavages along {110} in bcc iron emit dislocations from the curved crack fronts

Suzudo, Tomoaki; Ebihara, Kenichi; Tsuru, Tomohito; Mori, Hideki*

Scientific Reports (Internet), 12, p.19701_1 - 19701_10, 2022/11

https://doi.org/10.1038/s41598-022-24357-5

Body-centered-cubic (bcc) transition metals, such as -Fe and W, cleave along the {100} plane, even though the surface energy is the lowest along the {110} plane. To unravel the mechanism of this odd response, large-scale atomistic simulations of curved cleavage cracks of -Fe were conducted in association with stress intensity factor analyses of straight crack fronts using an interatomic potential created by an artificial neural network technique. The study provides novel findings: Dislocations are emitted from the crack fronts along the {110} cleavage plane, and this phenomenon explains why the {100} plane can be the cleavage plane. However, the simple straight crack-front analyses did not yield the same conclusion. It is suggested that atomistic modeling, at sufficiently large scales to capture the inherent complexities of materials using highly accurate potentials, is necessary to correctly predict the mechanical strength. The method adopted in this study is generally applicable to the cleavage problem of bcc transition metals and alloys.

Radiochemical research for the advancement of Mo/Tc generator by (n, ) method, 4

Fujita, Yoshitaka; Seki, Misaki; Ngo, M. C.*; Do, T. M. D.*; Hu, X.*; Yang, Y.*; Takeuchi, Tomoaki; Nakano, Hiroko; Fujihara, Yasuyuki*; Yoshinaga, Hisao*; et al.

KURNS Progress Report 2021, P. 118, 2022/07

https://www.rri.kyoto-u.ac.jp/PUB/report/PR/ProgRep2021/ProgressReport2021_online.pdf#page=134

no abstracts in English

Investigation on soundness of JMTR Facility piping by ultrasonic thickness measurement

Omori, Takazumi; Otsuka, Kaoru; Endo, Yasuichi; Takeuchi, Tomoaki; Ide, Hiroshi

JAEA-Review 2021-015, 57 Pages, 2021/11

JAEA-Review-2021-015.pdf:6.3MB

The JMTR reactor facility was selected as a decommissioning one in the Medium/Long-Term Management Plan of JAEA Facilities formulated on April 1, 2017. Therefore, the decommissioning plan was submitted to Nuclear Regulation Authority on September 18, 2019, and the approval was obtained on March 17, 2021 after two amendments. Currently, preparations for decommissioning are underway. The JMTR reactor facility has been aged for more than 50 years since the first criticality in March 1968. However, some of the water piping systems has not been updated since its construction, and there is a possibility of pipe wall thinning due to corrosion, etc. Therefore, the integrity of the water piping was investigated for the facilities that circulate cooling water and pump radioactive liquid waste. In this investigation, the main circulation system of the reactor primary cooling system, the pool canal circulation system, the CF pool circulation system, the drainage system of the liquid waste disposal system, and the hydraulic rabbit irradiation system of the main experimental facility were measured for the pipe wall thickness using the Ultrasonic Thickness Measurement (UTM) method. These values satisfied the technical standards for research and test reactor facilities. No loss of integrity is expected to occur during the upcoming decommissioning period. In the future, we will periodically confirm that there is no wall thinning in the piping of the cooling water circulation and the water transmission system during the decommissioning period by using this result as basic data.

Status of JMTR decommissioning plan formulation

Otsuka, kaoru; Hanakawa, Hiroki; Nagata, Hiroshi; Omori, Takazumi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko

UTNL-R-0496, p.13_1 - 13_11, 2018/03

no abstracts in English

Design of the pulse bending magnet for switching the painting area between the MLF and MR in J-PARC 3-GeV RCS

Takayanagi, Tomohiro; Yoshimoto, Masahiro; Watanabe, Masao; Saha, P. K.; Ueno, Tomoaki; Togashi, Tomohito; Yamazaki, Yoshio; Kinsho, Michikazu; Fujimori, Hiroshi*; Irie, Yoshiro*

Proceedings of 1st International Particle Accelerator Conference (IPAC '10) (Internet), p.3293 - 3295, 2010/05

http://accelconf.web.cern.ch/AccelConf/IPAC10/papers/wepd085.pdf

It has been designed the pulse bending magnet for switching the painting area to the MLF and MR, which is located in injection line area in J-PARC 3-GeV RCS. After upgrading the linac energy to 400 MeV, the two pulse magnets switch the injection beam orbit to change the each painting area by each shot of 25 Hz. The magnetic field strength switching by 25 Hz are required for the painting injection. Furthermore, in the case of the beam commissioning of the 3-GeV RCS, the center injection is performed. Therefore, the wide range of the beam bending angle from 3 mrad to 38 mrad is required for the pulse magnets.

Nuclear and thermal design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

Mori, Tomoaki; Nishihara, Tetsuo; Kunitomi, Kazuhiko

Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(3), p.253 - 261, 2007/09

https://doi.org/10.3327/taesj.J06.045

Design studies of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C), which can produce both electric power and hydrogen, have proceeded in Japan Atomic Energy Agency. In future, it will be obliged to operate using not enriched uranium but plutonium to coexist with fast reactors after the full deployment of fast reactor cycle. Therefore, a nuclear and thermal design has been performed to confirm the feasibility of the reactor core using Mixed-Oxide (MOX) fuel. The reactor core with operation period of 450 days and average burn-up of 123 GWd/ton for discharged fuel was designed. The reactor core met safety requirements of maximum fuel temperature of less than 1400 C during normal operation, maximum power density of less than 13 W/cm, shutdown margin of more than 1.0 %k/kk' and negative reactivity coefficient. The results proved that it is possible to operate the GTHTR300C using MOX fuels without consuming natural uranium resources.

JAEA's VHTR for Hydrogen and Electricity Cogeneration; GTHTR300C

Kunitomi, Kazuhiko; Yan, X.; Nishihara, Tetsuo; Sakaba, Nariaki; Mori, Tomoaki

Nuclear Engineering and Technology, 39(1), p.9 - 20, 2007/02

Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2020s. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of 950C. The maximum 370 MW to the secondary system is used for hydrogen production and the balance of the reactor thermal power is used for electricity generation. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the gas turbine, IHX, etc.

DELIGHT-8; One dimensional fuel cell burnup analysis code for High Temperature Gas-cooled Reactor (HTGR) (Joint research)

Nojiri, Naoki; Fujimoto, Nozomu; Mori, Tomoaki; Obata, Hiroyuki*

JAERI-Data/Code 2004-012, 65 Pages, 2004/10

JAERI-Data-Code-2004-012.pdf:7.77MB

DELIGHT code is a fuel cell burnup analysis code which can produce the group constants necessary for High Temperature Gas-cooled Reactors (HTGR) core analyses. Collision probability method is used to the lattice calculation. The lattice calculation model is a cylinder type fuel or a ball type fuel of the HTGR. This code characterizes the burnup calculation considering the double heterogeneity caused by coated fuel particles of the HTGR fuel. DELIGHT code has updated its nuclear data library to the latest JENDL-3.3 data, and included new burnup chain models in order to calculate high burnup HTGR cores. The material regions of the periphery burnable poisons (BPs) were divided into details in order to improve calculation accuracy of the BP lattice calculation. This report presents the revised points of the DELIGHT-8 and can be used as user's manual.

JASPER Experimental data book (VI); Special materials experiment

Mori, Tomoaki*; Takemura, Morio*

JNC TJ9450 2000-001, 96 Pages, 2000/03

JNC-TJ9450-2000-001.pdf:2.04MB

This report is intended to make it easier to apply the measured data obtained from the Special Materials Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about a month beginning at the end of June, 1992 as the last one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Special Materials Experiment was planned to obtain the data of neutron attenuation characteristics of selected shielding materials for use in advanced fast reactors. The material of particular interest for the experiment was zirconium hydride that is rich in hydrogen. The mockup slabs for the special materials were preceded by the spectrum modifier behind the TSR-II reactor of Tower Shielding Facility. The layer of zirconium hydride was simulated with a combination of zirconium and polyethylene slabs. The thick layer of polyethylene with no zirconium was installed in some configurations.Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured in eight configurations and neutron spectrum in high energy region was measured also in almost all configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12277. "Measurements for the JASPER Program Special Materials Experiment"). Additional information reported by the assignee is utilized also.