Application of analysis for assembly of integrated components to steel member connections for seismic safety assessment of plant structures, 2; Plastic analysis

Nishida, Akemi; Murakami, Takahiro*; Satoda, Akira*; Asano, Yuya*; Guo, Z. H.*; Oshima, Masami*; Matsukawa, Keisuke*; Nakajima, Norihiro

Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 10 Pages, 2022/07

no abstracts in English

Development of HCl-free solid phase extraction combined with ICP-MS/MS for rapid assessment of difficult-to-measure radionuclides, 1; Selective measurement of Zr and Mo in concrete rubble

Do, V. K.; Furuse, Takahiro; Murakami, Erina; Aita, Rena; Ota, Yuki; Sato, Soichi

Journal of Radioanalytical and Nuclear Chemistry, 327(1), p.543 - 553, 2021/01

https://doi.org/10.1007/s10967-020-07503-z

A new HCl-free chromatographic separation procedure has been developed for sequential separation of Zr and Mo from concrete matrices. Accordingly, Zr and Mo could be sensitively and selectively measured by ICP-MS/MS using ammonia reaction gas. The recoveries of greater than 90% for Zr and Mo from concretes could be achieved. The measurement condition was optimized for complete suppression of interferences from Nb and peak tailing from abundant isotopes of Zr and Mo in concrete matrices. The removal of interferences was verified by measurement of radio-contamination-free concretes used as a sample matrix blank. Method detection limits of 1.7 mBq g and 0.2 Bq g were achieved for Zr and Mo, respectively, in the concrete matrices. The interference removal factor for Nb (equivalent to the decontamination factor in radiochemical separation) was of the order of 10, and the abundance sensitivity was of the order of 10, indicating that the developed method is reliable for verifying the presence of ultralow concentrations of Zr and Mo. The present method is suitable for the rapid assessment of Zr and Mo for radioactivity inventory of concrete rubble.

Application of analysis for assembly of integrated components to steel member connections towards seismic safety assessment of plant structures

Nishida, Akemi; Murakami, Takahiro*; Satoda, Akira*; Asano, Yuya*; Guo, Z.*; Matsukawa, Keisuke*; Oshima, Masami*; Nakajima, Norihiro

Transactions of the 25th International Conference on Structural Mechanics in Reactor Technology (SMiRT-25) (USB Flash Drive), 10 Pages, 2019/08

Exhaustive studies on external events that may pose a threat to the structures of nuclear facilities and evaluations of the structural integrities are critical to safety. One of the components that greatly influence the behavior of the plant structure is the connection of structural members. In particular, the modeling of the connections has relied on empirical methods, and been conservatively designed and evaluated by considering them as pinned or rigid connections. Therefore, in this research, we have aimed to develop a connection modeling method that reproduces more realistic behavior by utilizing a three-dimensional model of the connection. As the first step of this research, we planned to determine the stiffness of the connections of steel structural members. The results confirmed that the connection can be regarded as a partially-restrained connection depending on the connection specifications of the structure, and the prospects for realistic stiffness evaluation of the connection were determined.

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*

Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09

https://doi.org/10.1080/00223131.2014.994049

A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.

Development of proposed guideline of flow-induced vibration evaluation for hot-leg piping in a sodium-cooled fast reactor

Sakai, Takaaki; Yamano, Hidemasa; Tanaka, Masaaki; Ono, Ayako; Ohshima, Hiroyuki; Kaneko, Tetsuya*; Hirota, Kazuo*; Sago, Hiromi*; Xu, Y.*; Iwamoto, Yukiharu*; et al.

Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 13 Pages, 2013/05

The development of flow-induced vibration evaluation methodology has reached a milestone that separate-effect experimental data under a high Reynolds number regime including swirl and deflected inflow conditions are available for the validation of the methodology. On the other hand, technical standards are desirable to be documented for designers of sodium-cooled fast reactors. From such a background, the documentation of a flow-induced vibration design guideline has been made for the hot-leg piping of Japan sodium-cooled fast reactor. This paper describes the design guideline of the flow-induced vibration evaluation methodology, which has been informed from main separate-effect experiments, as well as supplemental interpretation for the guideline.

Study on mechanism of pressure fluctuation in a primary circuit hot-leg pipings of a sodium-cooled fast reactor

Murakami, Takahiro*; Eguchi, Yuzuru*; Tanaka, Masaaki; Yamano, Hidemasa

Nihon Kikai Gakkai Rombunshu, B, 78(792), p.1388 - 1391, 2012/08

https://doi.org/10.1299/kikaib.78.1388

In a conceptual design of Japan Sodium-cooled Fast Reactor (JSFR), short-elbow pipings with large-diameter are adopted for cooling system, in which the coolant flow causes periodical pressure fluctuation at the short-elbow. However, the mechanism of the periodical pressure fluctuation has not been clarified yet. In this paper, unsteady three-dimensional analysis by a finite element large-eddy-simulation (LES) code is carried out in order to explore the mechanism of the periodical pressure fluctuation in short-elbow pipings, based on visualizations of dynamic flow structure in the numerical results.

Sulfate complexation of element 104, Rf, in HSO/HNO mixed solution

Li, Z.*; Toyoshima, Atsushi; Asai, Masato; Tsukada, Kazuaki; Sato, Tetsuya; Sato, Nozomi; Kikuchi, Takahiro; Nagame, Yuichiro; Schdel, M.; Pershina, V.*; et al.

Radiochimica Acta, 100(3), p.157 - 164, 2012/03

https://doi.org/10.1524/ract.2012.1898

Unsteady elbow pipe flow to develop a flow-induced vibration evaluation methodology for JSFR

Yamano, Hidemasa; Tanaka, Masaaki; Ono, Ayako; Murakami, Takahiro*; Iwamoto, Yukiharu*; Yuki, Kazuhisa*; Sago, Hiromi*; Hayakawa, Satoshi*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

http://www-pub.iaea.org/mtcd/meetings/PDFplus/2009/cn176/cn176_BoeS.pdf

This paper describes the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in JSFR, in particular emphasizing on recent R&D activities that investigate unsteady elbow pipe flow. The experiment using the 1/3-scale test section was performed to investigate the effect of swirl flow at the inlet. Although the flow separation region was distorted at the downstream from the elbow, the experiment clarified that the effect of swirl flow on pressure fluctuation onto the pipe wall was not significant. The simulation revealed that Reynolds number scarcely affects flow patterns and flow velocity distributions.

Experimental investigation and validation of neutral beam current drive for ITER through ITPA joint experiments

Suzuki, Takahiro; Akers, R.*; Gates, D. A.*; Gnter, S.*; Heidbrink, W. W.*; Hobirk, J.*; Luce, T. C.*; Murakami, Masanori*; Park, J. M.*; Turnyanskiy, M.*; et al.

Nuclear Fusion, 51(8), p.083020_1 - 083020_8, 2011/08

https://doi.org/10.1088/0029-5515/51/8/083020

Joint experiments investigating the off-axis neutral beam current drive (NBCD) capability to be utilized for advanced operation scenario development in ITER was conducted in 5 tokamaks (AUG, DIII-D, JT-60U, MAST and NSTX) through the ITPA. We discuss results obtained in the joint experiments, where the toroidal field, , covered 0.3-3.7 T, the plasma current, , 0.6-1.2 MA, and the beam energy, Eb, 67-350 keV. A current profile broadened by off-axis NBCD was observed in MAST. In DIII-D, good agreement between the measured and calculated NB driven current profile was observed. In JT-60U, agreement between measured and calculated NBCD location was obtained, when the NBCD location (0.3-0.6 in /), heating power (6-13 MW), triangularity (0.25-0.45), and (85 and 350 keV) were widely scanned. In AUG (at low 0.2) and DIII-D, introduction of a fast ion diffusion coefficient of 0.3-0.5 m/s in the calculation gave better agreement at high heating power (5 and 7.2 MW), suggesting anomalous transport of fast ions by turbulence. It was found through these ITPA joint experiments that NBCD related physics quantities reasonably agree with calculations (with = 0-0.5 m/s) in all devices when there is no MHD activity except ELMs. Proximity of measured off-axis beam driven current to the corresponding calculation with = 0 has been discussed for ITER in terms of a theoretically predicted scaling of fast-ion diffusion that depends on / for electrostatic turbulence or for electromagnetic turbulence.

Experimental investigation and validation of neutral beam current drive for ITER through ITPA joint experiments

Suzuki, Takahiro; Akers, R.*; Gates, D. A.*; Gnter, S.*; Heidbrink, W. W.*; Hobirk, J.*; Luce, T. C.*; Murakami, Masanori*; Park, J. M.*; Turnyanskiy, M.*; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2010/10

http://www-naweb.iaea.org/napc/physics/FEC/FEC2010/html/node57.htm#14693

Joint experiments investigating the off-axis neutral beam current drive (NBCD) capability to be utilized for advanced operation scenario development in ITER was conducted in 4 tokamaks (AUG, DIII-D, JT-60U and MAST) through the ITPA. We discuss results obtained in the joint experiments, where the toroidal field, Bt, covered 0.3-3.7 T, the plasma current, Ip, 0.6-1.2 MA, and the beam energy, Eb, 67-350 keV. A current profile broadened by off-axis NBCD was observed in MAST. In DIII-D, good agreement between the measured and calculated NB driven current profile was observed. In JT-60U, agreement between measured and calculated NBCD location was obtained, when the NBCD location (0.3-0.6 in r/a), heating power (6-13 MW), triangularity d (0.25-0.45), and Eb (85 and 350 keV) were widely scanned. In AUG (at low d 0.2) and DIII-D, introduction of a fast ion diffusion coefficient of Db 0.3-0.5 m/s in the calculation gave better agreement at high heating power (5 and 7.2 MW), suggesting anomalous transport of fast ions by turbulence. It was found through these ITPA joint experiments that NBCD related physics quantities reasonably agree with calculations (with Db=0-0.5 m/s) in all devices when there is no MHD activity except ELMs.