Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger; Project overview

Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.

Interfacial ferromagnetism and atomic structures in high-temperature grown FeO/Pt/FeO epitaxial trilayers

Kikkawa, Takashi*; Suzuki, Motohiro*; Ramos, R.*; Aguirre, M. H.*; Okabayashi, Jun*; Uchida, Kenichi*; Lucas, I.*; Anadn, A.*; Kikuchi, Daisuke*; Algarabel, P. A.*; et al.

Journal of Applied Physics, 126(14), p.143903_1 - 143903_11, 2019/10

https://doi.org/10.1063/1.5125761

Analysis of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya

Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02

https://doi.org/10.1080/00223131.2017.1384705

The criticality safety of the fuel debris in Fukushima Daiichi Nuclear Power Plant is one of the most important issues and the adoption of the burnup credit is desired for the criticality analysis. The assay data of used nuclear fuel irradiated in 2F2 is evaluated to validate SWAT4.0 for BWR fuel burnup problem. The calculation results revealed that number density of many heavy nuclides and FPs showed good agreement with the experimental data except for U, Np, Pu and Sm isotopes. The cause of the difference is assumption of the initial number density and void ratio and overestimation of the capture cross section of Np. The C/E-1 values do not depend on the types of fuel rods (UO or UO-GdO) and it is similar to that for the PWR fuel. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of the BWR fuel and it has sufficient accuracy to be adopted in the burnup credit evaluation of the fuel debris.

Enhancement of spin-Seebeck effect by inserting ultra-thin FeCu interlayer

Kikuchi, Daisuke*; Ishida, Masahiko*; Uchida, Kenichi*; Qiu, Z.*; Murakami, T.*; Saito, Eiji

Applied Physics Letters, 106(8), p.082401_1 - 082401_4, 2015/02

https://doi.org/10.1063/1.4913531

Generation of spin currents by surface plasmon resonance

Uchida, Kenichi*; Adachi, Hiroto; Kikuchi, Daisuke*; Ito, Shun*; Qiu, Z.*; Maekawa, Sadamichi; Saito, Eiji

Nature Communications (Internet), 6, p.5910_1 - 5910_8, 2015/01

https://doi.org/10.1038/ncomms6910

Spin-relaxation modulation and spin-pumping control by transverse spin-wave spin current in YFeO

Kajiwara, Yosuke*; Uchida, Kenichi*; Kikuchi, Daisuke*; An, Toshu*; Fujikawa, Yasunori*; Saito, Eiji

Applied Physics Letters, 103(5), p.052404_1 - 052404_4, 2013/07

https://doi.org/10.1063/1.4817076

Heat-current-induced manipulation of spin relaxation in YFeO under an in-plane temperature gradient is investigated. We show that the linewidth of the ferromagnetic resonance spectrum, i.e., the spin relaxation, in an YFeO film increases or decreases depending on the temperature-gradient direction and that this modulation is attributed to the spin-transfer torque caused by a thermally induced transverse spin-wave spin current in the YFeO film. The experimental results also show that the spin-current magnitude generated by spin pumping in an attached Pt film is inversely proportional to the square of the modulated Gilbert damping constant, consistent with a phenomenological spin-pumping model.

Development of 2012 edition of JSME code for design and construction of fast reactors, 6; Design margin assessment for the new materials to the rules

Ando, Masanori; Watanabe, Sota*; Kikuchi, Koichi*; Otani, Tomomi*; Sato, Kenichiro*; Tsukimori, Kazuyuki; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 11 Pages, 2013/07

New 2012 edition of JSME code for design and construction of fast reactors (FRs code) was published by Japan society of mechanical engineers (JSME). Main topic of the current JSME FRs code 2012 edition is registration of the two new materials, 316FR and Mod.9Cr-1Mo steel. The design margins for the new materials to the rules for the components and piping serviced at elevated temperature described in the JSME FRs code were assessed. To confirm the design margins, a series of the assessment program for the new materials to the conventional design rules was performed using the evaluation of the experimental data and finite element analysis. Through these assessments, the enough design margins for new materials to the rules were confirmed.

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2011 - March 31, 2012)

Katagiri, Hiromi; Okuno, Hiroshi; Okamoto, Akiko; Ikeda, Takeshi; Tamura, Kenichi; Nagakura, Tomohiro; Nakanishi, Chika; Yamamoto, Kazuya; Abe, Minako; Sato, Sohei; et al.

JAEA-Review 2012-033, 70 Pages, 2012/08

JAEA-Review-2012-033.pdf:6.38MB

When a nuclear emergency occurs in Japan, JAEA has the responsibility of providing technical support to the National government, local governments, etc., by the Basic Law on Emergency Preparedness and the Basic Plan for Disaster Countermeasures. NEAT of JAEA gives technical advice and information, dispatch specialists as required, supplies with the National Government and local governments emergency equipment and materials. NEAT provides various lectures and training courses concerning nuclear disaster prevention for emergency response organizations at normal time. Concerning the assistance to the Accident of Fukushima No.1 Nuclear Power Station caused by the Great East Japan Earthquake on 11 March, 2011, JAEA assisted activities including environmental radiation monitoring, environmental radioactivity analyses, resident public consulting etc., with its the utmost effort. This annual report summarized these activities of NEAT in the fiscal year 2011.

The Creep rupture strength evaluation in welded joint of Mod.9Cr-1Mo steel

Wakai, Takashi; Nagae, Yuji; Takaya, Shigeru; Obara, Satoshi; Date, Shingo*; Yamamoto, Kenji*; Kikuchi, Koichi*; Sato, Kenichiro*

Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku, 52(2), p.147 - 159, 2011/07

By employing high-Cr ferritic steels to the structural materials for JSFR, a compact plant designing can be achieved. It contributes to reduce the construction cost and to enhance the freedom of designing. Among the high-Cr ferritic steels, modified 9Cr-1Mo steel (compatible to ASTM A335 P91) is a candidate of the structural material for the demonstration facility of JSFR, because the steel has superior microstructure stability at elevated temperature for long time. However, remarkable creep strength degradation has been observed in the welded joint of high-Cr ferritic steels, especially in long-term region. It is known as "Type-IV damage". In the elevated temperature designing for the fast reactors, such creep strength degradation must be taken into account properly. Therefore, the creep strength assessment procedure and the allowable stress for the welded joints made of modified 9Cr-1Mo steel have been proposed. In this study, (1) a series of creep rupture tests to verify the validity of the creep strength assessment procedure was performed. (2) Applicability of the creep strength assessment procedure to the creep fatigue strength assessment of the welded joints made of modified 9Cr-1Mo steel was investigated. (3) Metallurgical examinations of creep ruptured specimens were carried out to confirm the contribution of "Type-IV damage". As a result, it was demonstrated that the creep strength assessment procedure was validated using the long-term creep rupture test results less than 30,000h and that the creep strength assessment procedure was applicable to the creep-fatigue strength assessment based on some uniaxial creep-fatigue test results.

AMATERAS; A Cold-neutron disk chopper spectrometer

Nakajima, Kenji; Kawamura, Seiko; Kikuchi, Tatsuya; Nakamura, Mitsutaka; Kajimoto, Ryoichi; Inamura, Yasuhiro; Takahashi, Nobuaki; Aizawa, Kazuya; Suzuya, Kentaro; Shibata, Kaoru; et al.

Journal of the Physical Society of Japan, 80(Suppl.B), p.SB028_1 - SB028_6, 2011/05

https://doi.org/10.1143/JPSJS.80SB.SB028

AMATERAS is a cold-neutron disk-chopper spectrometer in MLF, J-PARC. The construction of main part of the spectrometer has been completed in spring of 2009. Soon after that, we have started the commissioning work on AMATERAS. The performance of AMATERAS has been examined by test experiments in the course of commissioning. In parallel to these works, we have started the user program on AMATERAS from December 2009 and we are getting scientific results from our spectrometer. In this presentation, we will report the current status of AMATERAS including the results of performance tests and some of examples of scientific outputs.