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Journal Articles

Reaction behavior between sodium and molten salt caused by the heat transfer tube failure for sodium-cooled fast reactor coupled to thermal energy storage system

Sato, Rika; Kondo, Toshiki; Umeda, Ryota; Kikuchi, Shin; Yamano, Hidemasa

Progress in Nuclear Science and Technology (Internet), 8, p.137 - 142, 2025/09

https://doi.org/10.15669/pnst.8.137

In a sodium-cooled fast reactor (SFR) coupled to thermal energy storage (TES) system, the reaction between nitrate molten salt as thermal energy storage medium and sodium (Na) as reactor coolant might occur under postulated accidental conditions. Thus, the reaction behavior of Na-nitrate molten salt is one of the important phenomena in terms of safety assessment of the SFR with TES system. In this study, reaction experiments on Na-solar salt were performed. It was found that Na-solar salt reaction occurred after the NaNO $_{3}$ -KNO $_{3}$ eutectic melting. Based on the measured reaction temperature, the kinetic parameters and rate constant were obtained and compared with the sodium-water reaction. From the results of kinetic analysis, it could be assumed that Na-solar salt reaction occurs in the time frame of the accident such as the failure of heat transfer tube of sodium-molten salt heat exchanger.

JAEA Reports

Analysis of behavior of Ru with nitrogen oxide chemical behavior in accident of evaporation to dryness by boiling of reprocessed high level liquid waste

Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*

JAEA-Research 2021-005, 25 Pages, 2021/08

JAEA-Research-2021-005.pdf:2.91MB

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an empirical correlation equation of Ru mass transfer coefficient across the vapor-liquid surface, which can be useful for quantitative simulation of Ru mitigating behavior, has been obtained from data analyses of small-scale experiments conducted to clarify gaseous Ru migrating behavior under steam-condensing condition. A simulation study has been also carried out with a hypothetical typical facility building successfully to demonstrate the feasibility of quantitative estimation of amount of Ru migrating in the facility using the obtained correlation equation implemented in SCHERN computer code which simulates chemical behaviors of nitrogen oxide based on the condition also simulated thermal-hydraulic computer code.

Journal Articles

Development of an integrated computer code system for analyzing irradiation behaviors of a fast reactor fuel

Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi

Nuclear Technology, 207(8), p.1280 - 1289, 2021/08

AA2020-0247.pdf:1.83MB

https://doi.org/10.1080/00295450.2020.1810977

Times Cited Count：3 Percentile：24.57(Nuclear Science & Technology)

Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.

Journal Articles

Numerical investigation on unstable behaviors of cellular premixed flames at low Lewis numbers based on the diffusive-thermal model and compressible Navier-Stokes equations

Thwe Thwe, A.; Kadowaki, Satoshi; Hino, Ryutaro

Journal of Thermal Science and Technology (Internet), 13(2), p.18-00457_1 - 18-00457_12, 2018/12

https://doi.org/10.1299/jtst.2018jtst0034

Times Cited Count：0 Percentile：0.00(Thermodynamics)

Two dimensional unsteady calculations of reactive flows were performed in large domain to investigate the unstable behaviors of cellular premixed flames at low Lewis numbers based on the diffusive-thermal (D-T) model and compressible Navier-Stokes (N-S) equations. The growth rates obtained by the compressible N-S equations were large and the unstable ranges were wide compared with those obtained by the D-T model equations. When the length of computational domain increased, the number of small cells separated from large cells of the cellular flame increased drastically. The stronger unstable behaviors and the larger average burning velocities were observed especially in the numerical results based on the compressible N-S equations. In addition, the fractal dimension obtained by the compressible N-S equations was larger than that by the D-T model equations. Moreover, we confirmed that the radiative heat loss promoted the instability of premixed flames at low Lewis numbers.

Journal Articles

Thermal-hydraulic safety research on nuclear containment vessel at JAEA

Shibamoto, Yasuteru; Yonomoto, Taisuke; Hotta, Akitoshi*

Nihon Genshiryoku Gakkai-Shi ATOMO, 58(9), p.553 - 557, 2016/09

no abstracts in English

Journal Articles

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, 1; Thermal mixing behavior of helium gas in HTTR

Tochio, Daisuke; Fujimoto, Nozomu

Journal of Nuclear Science and Technology, 53(3), p.425 - 431, 2016/03

https://doi.org/10.1080/00223131.2015.1054910

Times Cited Count：1 Percentile：9.13(Nuclear Science & Technology)

The future HTGR is now designed in JAEA. The reactor has many merging points of helium gas with different temperature. It is needed to clear the mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the HTTR due to lack of mixing of helium gas in the primary cooling system. Now the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal-hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the mixing behavior of helium gas. As the result, it was confirmed that the mixing behavior of helium gas in the primary cooling system is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.

Journal Articles

Recent results from LOCA study at JAERI

Nagase, Fumihisa; Fuketa, Toyoshi

NUREG/CP-0185, p.321 - 331, 2004/00

With a view to obtaining basic data to evaluate high burnup fuel behavior under loss of coolant accident (LOCA) conditions, a research program is being conducted at the Japan Atomic Energy Research Institute (JAERI). The program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. Hydrogen effects have been especially examined because hydrogen absorption has the great impact on cladding embrittlement. The tests on irradiated claddings have recently been started and preliminary results have been obtained. The present paper summarizes recent results from those studies.

JAEA Reports