Aerosol experiments and measurement techniques on pool scrubbing-related source term issues

Lebel, L. S.*; Morreale, A. C.*; Freitag, M.*; Gupta, S.*; Allelein, H.-J.*; Klauck, M.*; Sun, Haomin; Herranz, L. E.*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

A Comparative CFD exercise on bubble hydrodynamics using Euler-Euler and interface tracking approaches

Dehbi, A.*; Cheng, X.*; Liao, Y.*; Okagaki, Yuria; Pellegrini, M.*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 15 Pages, 2022/03

Development of numerical analysis codes for multi-level and multi-physics approaches in an advanced reactor design study

Tanaka, Masaaki; Doda, Norihiro; Mori, Takero; Yokoyama, Kenji; Uwaba, Tomoyuki; Okajima, Satoshi; Matsushita, Kentaro; Hashidate, Ryuta; Yada, Hiroki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Japan Atomic Energy Agency is developing an innovative design system named ARKADIA to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In the first phase of its development, ARKADIA-Design for design study and ARKADIA-Safety for safety assessment will be developed individually. In this paper, focusing on the ARKADIA-Design, the concept of the system is described and numerical analysis codes to be used for the multi-level and multi-physics analyses are introduced. Descriptions of the practical functions composed by the analysis codes and the representative problems in application studies for validation are introduced.

Preliminary application of eutectic reaction model on boron carbide and stainless steel to severe accident simulation of sodium-cooled fast reactors

Yamano, Hidemasa; Morita, Koji*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

For a severe accident (SA) simulation of sodium-cooled fast reactors, a eutectic reaction model between boron carbide (BC) and stainless steel (SS) has been developed to be incorporated into the SA simulation codes: SIMMER-III/IV. To confirm the applicability of SIMMER-IV involving the eutectic reaction model to reactor simulations, this study has preliminarily applied this code with the newly developed physical model to a SA simulation of a large-scale SFR designed in Japan. The simulation results show that the eutectic reaction is caused by the contact between the liquid SS and the broken BC pellets which are released to the coolant channel after the failure of cladding which is melted by the mixture of liquid SS and fuel particles coming from the neighboring fuel assemblies. The liquid eutectic material formed by the reaction stayed in the control assembly and the neighboring fuel assemblies. This preliminary simulation shows that the spreading area of BC-SS eutectic formation is limited within this calculation time.

Development of reacting jet evaluation model based on engineering approaches with particle method for improvement of LEAP-III code

Kosaka, Wataru; Uchibori, Akihiro; Takata, Takashi; Yanagisawa, Hideki*; Watanabe, Akira*; Jang, S.*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 11 Pages, 2022/03

For the safety assessment of a steam generator (SG) in a sodium-cooled fast reactor, the analysis code LEAP-III can evaluate the water leak rate during the long-term event progress including the tube failure propagation triggered by an occurrence of a small water leak in a failed heat transfer tube in SG. The LEAP-III has the advantage in completing the calculation with low computational cost since it consists of semi-empirical formulae and one-dimensional equations of conservation. However, an evaluation model of temperature distribution by the reacting jet provides wider high temperature region than the experimental data. As a result, LEAP-III shows excessive conservativeness in some case. A Lagrangian particle method code based on engineering approaches has been developed in order to improve this model to get more realistic temperature distribution. In this method, the jet behavior and chemical reaction are simulated using Newton's equation of motion with several engineering approximations instead of solving multi-dimension multiphase thermal hydraulic equations with sodium-water reaction. In this study, interparticle interaction force model was added, and also the chemical reaction and gas-liquid heat transfer evaluation models were improved. We conducted a test analysis, and compared the results by this particle method with the ones by SERAPHIM, that is a mechanistic analysis code for multi-dimensional multiphase flow considering compressibility and sodium-water reaction. Through this test analysis, it confirmed that this particle method has the basic capability to get a realistic temperature distribution with low computational cost, and also to predict tube failure occurrence by coupled with LEAP-III.

Numerical simulation of two-phase flow in fuel assemblies with a spacer grid using a mechanistically based method

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.

Integration of pool scrubbing research to enhance source-term calculations (IPRESCA) project

Gupta, S.*; Herranz, L. E.*; Lebel, L. S.*; Sonnenkalb, M.*; Pellegrini, M.*; Marchetto, C.*; Maruyama, Yu; Dehbi, A.*; Suckow, D.*; Krkel, T.*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Application of noncontact type of ultrasonic flowmeter for high temperature LBE flow

Obayashi, Hironari; Yamaki, Kenichi*; Kita, Satoshi*; Ariyoshi, Gen; Sasa, Toshinobu

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 11 Pages, 2022/03

JAEA developed a non-contact type flowmeter that two ultrasonic sensors were installed on outer surface of a cylindrical-shaped test section. The considerable issue of the realization was multiple reflection of emitted signal in the piping. Furthermore, wettability between LBE and inner surface of the test section was one of the important issue on the ultrasonic technique. To mitigate the multiple reflection, a critical angle as an incident angle not to produce a longitudinal wave on the solid / solid boundary was applied. And fine mirror-finished treatment was applied to the inner surface of the test section to ensure wettability. As a result of experiment at 350 degrees C, the observed correlation value between the non-contact type and the established plug type was 0.9986. Additionally, the developed non-contact type provided its sufficiently stable output during a long-term test more than 1,500 hours.

Unstructured-mesh simulation of sodium-water reaction in tube bundle system by SERAPHIM code

Uchibori, Akihiro; Shiina, Yoshimi*; Watanabe, Akira*; Takata, Takashi*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 12 Pages, 2022/03

An unstructured mesh-based analysis method has been integrated into the sodium-water reaction analysis code, SERAPHIM, in our recent studies. In this study, numerical analysis of an experiment on sodium-water reaction in a tube bundle domain was performed to investigate the effect of the unstructured mesh. The unrealistic behavior appeared in the coarse structured mesh was improved by the unstructured mesh. The numerical result in the case of the unstructured mesh reproduced the peak value of the temperature in the reacting flow.

Development of the sodium pool and floor concrete module for the integrated SFR safety analysis code, SPECTRA

Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03