Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 1; Proposal of countermeasures to mitigate temperature fluctuations around control rods

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.89 - 96, 2021/10

Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.

Alternation of magnetic anisotropy accompanied by metal-insulator transition in strained ultrathin manganite heterostructures

Kobayashi, Masaki*; Anh, L. D.*; Suzuki, Masahiro*; Kaneta-Takada, Shingo*; Takeda, Yukiharu; Fujimori, Shinichi; Shibata, Goro*; Tanaka, Arata*; Tanaka, Masaaki*; Oya, Shinobu*; et al.

Physical Review Applied (Internet), 15(6), p.064019_1 - 064019_10, 2021/06

https://doi.org/10.1103/PhysRevApplied.15.064019

Direct observation of the magnetic ordering process in the ferromagnetic semiconductor GaMnAs via soft X-ray magnetic circular dichroism

Takeda, Yukiharu; Oya, Shinobu*; Pham, N. H.*; Kobayashi, Masaki*; Saito, Yuji; Yamagami, Hiroshi; Tanaka, Masaaki*; Fujimori, Atsushi*

Journal of Applied Physics, 128(21), p.213902_1 - 213902_11, 2020/12

https://doi.org/10.1063/5.0031605

Study on multi-dimensional core cooling behavior of sodium-cooled fast reactors under DRACS operating conditions

Ezure, Toshiki; Onojima, Takamitsu; Tanaka, Masaaki; Kobayashi, Jun; Kurihara, Akikazu; Kameyama, Yuri*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.3355 - 3363, 2019/08

Steady-state sodium experiments under the operating conditions of a decay heat removal system (DHRS) were carried out as part of the safety enhancement of sodium-cooled fast reactors using the PLANDTL 2 facility, which has 30 heated channels with electric heaters and 25 no-heated channels as the simulated core. In the experiments, a direct reactor auxiliary cooling system (DRACS) with a dipped type direct heat exchanger (DHX) in the upper plenum was used as the DHRS. This paper reports on the preliminary experimental results of the PLANDTL 2 experiments under the DRACS operating conditions without flow in the primary circuit, including the thermal hydraulic interactions between the upper plenum and the core under the DHX operating conditions and the resulting core cooling behavior from the outside of the multiple rows of the fuel assemblies

Current status and future developments of hard X-ray photoelectron spectroscopy

Kobata, Masaaki; Okane, Tetsuo; Kobayashi, Keisuke*

Bunko Kenkyu, 67(4), p.161 - 162, 2018/08

We introduce hard X-ray photoelectron spectroscopy, which has been rapidly introduced and developed in synchrotron radiation facilities. In particular, in order to realize electronic state analysis by hard X-ray photoelectron spectroscopy of insulators, the developed charge neutralization method was described. As an example, we showed adsorption behavior of cesium to nuclear reactor structure assuming Fukushima Daiichi Nuclear Power Plant accident. Finally, future prospects of hard X-ray photoelectron spectroscopy will be described.

Evaluation of sediment and Cs redistribution in the Oginosawa River catchment near the Fukushima Dai-ichi Nuclear Power Plant using integrated watershed modeling

Sakuma, Kazuyuki; Malins, A.; Funaki, Hironori; Kurikami, Hiroshi; Niizato, Tadafumi; Nakanishi, Takahiro; Mori, Koji*; Tada, Kazuhiro*; Kobayashi, Takamaru*; Kitamura, Akihiro; et al.

Journal of Environmental Radioactivity, 182, p.44 - 51, 2018/02

https://doi.org/10.1016/j.jenvrad.2017.11.021

The Oginosawa River catchment lies 15 km south-west of the Fukushima Dai-ichi nuclear plant. The General-purpose Terrestrial Fluid-flow Simulator (GETFLOWS) code was used to study sediment and Cs redistribution within the catchment. Cesium-137 input to watercourses came predominantly from land adjacent to river channels and forest gullies. Forested areas far from the channels only made a minor contribution to Cs input to watercourses, total erosion of between 0.001-0.1 mm from May 2011 to December 2015. The 2.3-6.9% y decrease in the amount of Cs in forest topsoil over the study period can be explained by radioactive decay (approximately 2.3% y), along with a migration downwards into subsoil and a small amount of export. The amount of Cs available for release from land adjacent to rivers is expected to be lower in future than compared to this study period, as the simulations indicate a high depletion of inventory from these areas.

In-beam -ray spectroscopy of Mg via knockout reactions at intermediate energies

Momiyama, Satoru*; Doornenbal, P.*; Scheit, H.*; Takeuchi, Satoshi*; Niikura, Megumi*; Aoi, Nori*; Li, K.*; Matsushita, Masafumi*; Steppenbeck, D.*; Wang, H.*; et al.

Physical Review C, 96(3), p.034328_1 - 034328_8, 2017/09

https://doi.org/10.1103/PhysRevC.96.034328

no abstracts in English

Influence of inlet velocity condition on unsteady flow characteristics in piping with a short elbow under a high-Reynolds-number condition

Ono, Ayako; Tanaka, Masaaki; Kobayashi, Jun; Kamide, Hideki

Mechanical Engineering Journal (Internet), 4(1), p.16-00217_1 - 16-00217_15, 2017/02

http://doi.org/10.1299/mej.16-00217

In the design of the Advanced Sodium-cooled Fast Reactor in Japan, the Reynolds number in the primary hot leg (H/L) piping reaches 4.210. Furthermore, a short elbow is used in the H/L piping to achieve a compact plant layout. In the H/L piping, flow-induced vibration is a concern due to the excitation force caused by pressure fluctuation in the short elbow. In this report, the influence of inlet velocity condition on the unsteady velocity characteristics in the short elbow was studied by controlling the flow patterns at the elbow inlet. Measured velocity distributions indicated that the inlet velocity profiles affected a circumferential secondary flow, which then affected an area of flow separation at the elbow. It was also found that the velocity fluctuation at low frequency components observed upstream of the elbow could remain in downstream of the elbow though its intensity was attenuated.

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor; Influence of flow collector of backup CR guide tube

Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. Meanwhile, a self-actuated shutdown system (SASS) is equipped in a backup control rod (BCR) channel to ensure reactor shutdown. The BCR guide tubes have a flow guide structure "flow-collector" to provide reliable operation of SASS. Flow-collector may affect the thermal mixing behavior at the bottom of the UIS. This study has investigated the influence of the flow- collector on characteristics of the temperature fluctuation around the BCR channels.

Fundamental validation of fluid-structure thermal interaction simulation code for thermal striping in sodium-cooled fast reactors with parallel triple jets mixing experiments

Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*

Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-6) (Internet), 12 Pages, 2016/09

A numerical simulation code named MUGTHES which can deal with conjugate heat transfer problem between the fluid and the structure parts has been developed in order to predict the thermal response in the structure for estimation of the thermal fatigue issue. To perform fundamental validation of the MUGTHES, the benchmark simulation was considered using the experiment of planar triple parallel jets mixing sodium test (PLAJEST). Since it was known by literatures that three representative flow mixing patterns were shown in accordance with the velocity rate of the side jets to the center jet, three typical experimental conditions in the PLAJEST were employed as boundary conditions for the benchmark. Through the numerical simulations, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model to simulate thermal striping phenomena was confirmed.