New JENDL-4.0/HE neutron and proton ACE files

Konno, Chikara

Journal of Nuclear Science and Technology, 6 Pages, 2023/00

https://doi.org/10.1080/00223131.2023.2237970

The JENDL-4.0/HE neutron and proton ACE files were produced in 2017 and those of 22 nuclei for neutron and 25 nuclei for proton were bundled in the PHITS code. Recently it was found that the following five data in the JENDL-4.0/HE neutron and proton ACE files had any problems; ACE files for N and O, heating numbers, damage energy production cross sections, secondary neutron multiplicities and fission cross sections. Thus new JENDL-4.0/HE neutron and proton ACE files were produced with the problems fixed. This paper describes the problems and how to produce the new neutron and proton ACE files in detail.

Direct temperature measurement for lamp-based heating device

Sumita, Takehiro; Sudo, Ayako; Takano, Masahide; Ikeda, Atsushi

Science and Technology of Advanced Materials; Methods (Internet), 2(1), p.50 - 54, 2022/02

https://doi.org/10.1080/27660400.2022.2035097

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04

https://doi.org/10.1115/1.4045450

Measurement of local temperature around the impact points of fast ions under grazing incidence

Kokabu, Hiroki*; Yoon, S.*; Lee, H.*; Nakajima, Kaoru*; Matsuda, Makoto; Sataka, Masao*; Tsujimoto, Masahiko*; Toulemonde, M.*; Kimura, Kenji*

Nuclear Instruments and Methods in Physics Research B, 460, p.34 - 37, 2019/12

https://doi.org/10.1016/j.nimb.2019.01.006

Development of numerical analysis method for tube failure propagation under sodium-water reaction accident

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

https://asme.pinetec.com/icone26/

Development of non-transfer type plasma heating technology to address CMR behavior during severe accident with BWR design conditions

Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

https://doi.org/10.1115/1.4032478

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Computational and experimental examination of simulated core damage and relocation dynamics of a BWR fuel assembly

Hanus, G.*; Sato, Ikken; Iwama, Tatsuya*

Proceedings of International Waste Management Symposia 2016 (WM2016) (Internet), 12 Pages, 2016/03

JAEA plans a large-scale test to evaluate damage and relocation behavior of BWR core materials consisting of fuel rods, channel boxes, control blade and lower support structures. Its purpose is to contribute to understanding of core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing, JAEA desires preliminary investigations to examine melting test pieces. The purpose of such tests is to verify the materials and test piece will be heated by plasma to the target temperature (ca.2900K) and to collect data about the material relocation behavior. Results from preliminary computational simulations are presented illustrating the effectiveness of a 150 kW non-transferred plasma jet. An experimental test program using the computational analyses as a basis and a plasma torch is described.

Influence of heating method on size and morphology of metallic oxide powder synthesized from metallic nitrate solution

Segawa, Tomoomi; Fukasawa, Tomonori*; Yamada, Yoshikazu; Suzuki, Masahiro; Yoshida, Hideto*; Fukui, Kunihiro*

Proceedings of Asian Pacific Confederation of Chemical Engineering 2015 (APCChE 2015), 8 Pages, 2015/09

https://search.informit.com.au/browsePublication;isbn=9781922107473;res=IELENG#241

A mixed solution of uranyl nitrate and plutonium nitrate is converted to MOX raw powder by the microwave heating de-nitration method in nuclear reprocessing. Copper oxide synthesized by heating de-nitration was used as a model for the de-nitration process. The microwave heating method (MW) and infrared heating method (IR) were used, and how they and their heating rate influence the obtained particle morphology and size were investigated. The particles obtained by the MW and IR were sufficiently similar in the surface morphology and the mass median diameter was decreased by the increased heating rate. The mass median diameters by the MW were the heating rate and smaller than those obtained by IR. The particle size distribution of the particle obtained by the MW was broader than that by the IR. The relationship of the temperature distribution and particle size distribution by the MW was discussed by the numerical simulation.