Experimental investigation of debris sedimentation behaviour on bed formation characteristics

Shamsuzzaman, M.*; Horie, Tatsuro*; Fuke, Fusata*; Kai, Takayuki*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*; Tagami, Hirotaka; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12

Investigation on sedimentation behavior of debris is important to evaluate the sequence of core disruptive accident in SFR. To clarify this behavior, a series of experiments was performed by gravity driven discharge of solid debris from a nozzle into a water pool. The discharged debris accumulates on the collector plate at the bottom, forming either a Gaussian-type convex or ring-type concave mound depending on the experiment parameters. Three types of spherical debris with three diameters are employed to study the effect of experiment parameters on mound height of debris bed. During the experiment, mound height becomes decreasing with nozzle diameter and increasing with debris volume, which exhibits descending tendency in asymmetrical fashion with density variation and an unalike variation in mound height was observed with debris diameter. An empirical model was developed applying dimensional analysis to predict the variation in mound height of debris bed during sedimentation process.

Numerical simulation of melt-down behavior in SFR severe accidents by the MUTRAN code

Kubota, Ryuzaburo*; Yamada, Yumi*; Koyama, Kazuya*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kubo, Shigenobu; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12

This paper describes a melt-down event progression revealed by a numerical simulation in the protected loss of heat sink (PLOHS) event for Japan Sodium-cooled Fast Reactor (JSFR). A multi-component multi-field computer code, MUTRAN, has been applied in order to simulate complicated core material motions and associated heat-transfer phenomena among the materials in a degradation core. The analyses with MUTRAN covered core degradation behaviors from the intact geometry and addressed the two initial states: one was the core without the coolant as the leakage type, and the other was the core covered by the coolant only up to the top of the fissile fuel as the boiling type. The analyses revealed representative event progression.

Estimation of component failure rates for PSA in sodium-cooled fast reactor

Naruto, Kenichi*; Kurisaka, Kenichi

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12

Sodium experiment on fully natural circulation systems for decay heat removal in Japan Sodium-cooled Fast Reactor

Ono, Ayako; Kobayashi, Jun; Kamide, Hideki; Watanabe, Osamu*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor. The DHRS of JSFR consists of one unit of DRACS and two units of PRACS. In this study, the sodium experiments were conducted using a sodium test loop PLANDTL in order to investigate the effect of operation mode on transient behavior of thermal hydraulic in PRACS loop. The experimental results revealed the effect of increasing heat removal capacity of PRACS and the forced flow operation in PRACS loop on the thermal transient and natural circulation behavior in PRACS loop.

Numerical simulation of bubbling fluidized beds by coupling multi-fluid model with discrete element method

Guo, L.*; Morita, Koji*; Tobita, Yoshiharu

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12

In the safety analysis of liquid-metal fast reactors, thermal-hydraulic phenomena of multicomponent, multiphase flows in core disruptive accidents are regarded as particular difficulties. Accurate prediction of dispersed particle behaviors in such complicate flows is one of the key issues to be solved in numerical simulations. On the other hand, bubbling fluidization of particle beds is not only considered as an essential phenomenon in some industry areas, but also employed to understand the particle behaviors in the research field. In this study, a hybrid method for numerical simulations of bubbling fluidized beds was developed by combining the discrete element method with the multi-fluid model. A typical system of bubbling fluidized beds with glass particles is analyzed to validate the developed coupling algorithm. It was indicated that the present models and methods could provide a useful means for the numerical simulation of bubbling fluidization phenomena in particle beds.

Experimental study on material relocation during core disruptive accident in sodium-cooled fast reactors; Results of a series of fragmentation tests for molten oxide discharged into a sodium pool

Matsuba, Kenichi; Kamiyama, Kenji; Konishi, Kensuke; Toyooka, Junichi; Sato, Ikken; Zuev, V. A.*; Kolodeshnikov, A. A.*; Vasilyev, Y. S.*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12

A series of fragmentation tests (FR tests) for molten oxide was conducted to obtain experimental knowledge on the distance for fragmentation of molten core material discharged into the lower sodium plenum. Approx. 714 kg of molten alumina was discharged into a sodium pool (depth: 1.3 m, diameter: 0.4 m, temperature: approx. 673 K) through a duct (inner diameter: 4063 mm). The test results showed that the molten alumina was fragmented into particles within 1 m from the surface of the sodium pool. The estimated distances for fragmentation in the FR tests were roughly 6070% lower than the predictions by the existing representative correlation. The experimental knowledge confirms the possibility that the distance for fragmentation of the molten core material can be significantly reduced due to thermal interactions in the lower sodium plenum.

Validation of the SIMMER-IV severe accident computer code on three-dimensional sloshing behavior

Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Matsumoto, Tatsuya*; Morita, Koji*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12

In this study, numerical calculations were carried out, providing that the fluid-dynamics model of SIMMER-IV was valid to simulate the sloshing behavior. Comparing to the conventional two-dimensional simulation, it was found that the three-dimensional simulation can mitigate the fuel compaction to the center because the effect of circumferential momentum dissipation can be addressed. From these calculations, the validity of the SIMMER-IV code was confirmed for the sloshing behavior.

Investigation of dominant factors for evaluation of sodium leak and fire accident consequences by sensitivity analyses

Ohno, Shuji; Hamase, Erina; Kamide, Hideki

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 5 Pages, 2012/12

Sensitivity analyses of sodium leak and fire are performed to identify the dominant factors for the accident consequences. The analyses with a multi-cell zone model code SPHINCS treat sodium spray and pool simultaneous combustion and heat-mass transfer behaviors in a large-scale two-cell geometry. Atmospheric gas pressure increase and temperature increase of floor steel plate below the sodium pool are analyzed as the figures of merit to be directly focused on. The analyses clarify the important and dominant factors of the phenomena in the accident sequence quantitatively, resulting in the acquirement of the knowledge to conduct the appropriate code validation activity and to discuss the uncertainty in the safety evaluation results.

Numerical prediction of two-phase flow behavior under seismic condition by use of interface tracking method

Yoshida, Hiroyuki; Nagatake, Taku; Takase, Kazuyuki; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12

Measurements of surface heat flux and surface temperature distributions under a nucleate pool boiling bubble under low pressure

Liu, W.; Nagatake, Taku; Takase, Kazuyuki

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12

Nucleate boiling has high heat transfer coefficient and is widely used in heat exchangers, boilers and BWR, etc. The nucleate boiling has been studied extensively but its heat transfer mechanism is still unknown. Recently, JAEA has developed a technology for the measurements of the surface heat flux and surface temperature distributions. The technology solved inverse heat conduction problem to get the surface heat flux and surface temperature distributions, with using a measured inside wall temperature distribution data from self-developed high sensitive temperature sensors. The temperature sensors are allocated in high density at a depth several micro meters beneath the heated surface. In this paper, with using the developed technology, we measured the surface heat Flux and surface temperature distributions under a nucleate pool boiling bubble at both atmospheric pressure and a lower pressure. To control the position of boiling bubble, we mirror processed the heated surface and formed a nucleus at the center of the surface. Temperature sensors are allocated beside the nucleus, in a density of upmost 6 points per 1mm and at a depth about 1.4 m beneath the surface. Based on the experimental data, we examined boiling heat transfer models.