Investigation on flow field around a slit of Upper Internal Structure (UIS) in a highly compact vessel of a sodium cooled fast reactor

Kamide, Hideki; Aizawa, Kosuke; Oshima, Jun*; Nakayama, Okatsu*; Kasahara, Naoto

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11

Development of advanced loop type sodium cooled fast reactor is under going. An upper internal structure (UIS) has a radial slit to reduce the reactor vessel diameter. This UIS slit allows a high velocity from the core fuel subassemblies and influences the gas entrainment in the reactor vessel and also the delayed neutron precursor sampling for a failed fuel detection and location system. Then flow visualization and velocity measurements were carried out in an 1/10 scale water test model. The velocity measurement using particle image velocimetry showed that velocity in the slit region was accelerated at the heights of the UIS horizontal plates and kept higher value at the middle height of the upper plenum. Numerical simulation using a commercial CFD code was also carried out for this complex geometry of UIS to know adequate simulation method. The comparisons of velocity profiles in the UIS between the experiment and analysis showed good agreements.

Study of hydraulic behavior for reactor upper plenum in sodium-cooled fast reactor; Verification analysis of water experiment and applicability of vortex prediction method

Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru; Kamide, Hideki; Kimura, Nobuyuki; Nakayama, Okatsu; Ohshima, Hiroyuki; Narita, Hitoshi*; Fujimata, Kazuhiro*; Itooka, Satoshi*

JAEA-Research 2006-017, 113 Pages, 2006/03

JAEA-Research-2006-017.pdf:14.98MB

A conceptual design study of the sodium-cooled fast reactor is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems. Reduced scale water experiments are being performed in order to clarify the flow pattern in the upper plenum of the reactor which has higher velocity condition than the past design. In this report, the hydraulic analyses of the water experiments using the general-purpose thermal hydraulic analysis program were executed; and the applicability to evaluation of flow pattern and vortex cavitations for the designed reactor was examined. (1) Steady-state analyses under the Froude number similar condition were carried out for the 1/10th reduced scale plenum experiments. Analyses results reproduced the characteristic flow patterns in the upper plenum, such as gushed flow from the inside of the upper internal structure to reactor vessel wall and the jet flow from the slit of the upper internal structure. Further, it was confirmed that the calculated flow pattern of a designed reactor system agreed with that of the water experiment qualitatively. Moreover, the influence which setting of numerical solution and boundary condition etc. in analyzing causes to flow pattern in the plenum became clear. (2) The distribution of the vortices under the dipped plate region in the 1/10th plenum model was evaluated using the prediction method of a submerged vortex which is based on the stretching vortex theory. In case of the same velocity condition as the reactor, it identified the two vortices which were sucked into the hot leg piping from the cold leg piping wall as the submerged vortex cavitations. From this analysis result, it confirmed that the submerged vortex cavitations, which may occur in the reactor upper plenum steadily, could be identified using this prediction method.

Hydraulic Experiment on Flow Optimization in Compact Reactor Vessel; Effects of double dipped plates and outer shell of UIS

Nakayama, Oukatsu; Kimura, Nobuyuki; Kamide, Hideki; Ito, Masami*; Sekine, Tadashi*

JNC TN9400 2005-032, 103 Pages, 2005/03

JNC-TN9400-2005-032.pdf:33.44MB

On the FBR feasibility study, an innovative sodium cooled fast reactor has been investigated in Japan Nuclear Cycle Development Institute. In order to reduce the construction cost of reactor, the compact reactor vessel is designed. Gas entrainment at the free surface is one of issues to be solved because of high velocity in the upper plenum vessel. We performed an 1/10th scaled model water experiment for the upper plenum of reactor vessel and investigated flow fields in the plenum in order to optimize flow. In the last experiment, we established flow optimization structures for reduction of flow toward to the free surface and restraint on the vortex cavitations near H/L inlet, i.e., plug into the dipped plate hole to install a fuel handling mechanism (FHM) and a flow splitter at vessel wall. In this report, effects of double dipped plates (D/Ps) and outer shell of upper inner structure (UIS) are investigated. In the result, we could confirm that the double D/P structure set below the free surface is effective to reduce flow toward the free surface between reactor vessel wall and guide tube of fuel transfer machine. An FHM plug, which has a step structure at mid height between double D/Ps, showed that it reduced flow velocity through the gap between the plug and the D/P. A perforated outer shell was set around the UIS to reduce vortex strength developed from the cold legs to the hot leg intake and also to restrain the vortex cavitation. It was shown that the onset condition of vortex cavitation at reactor vessel wall was reduced and the cavitation occurred easily. At the cold leg the onset condition of the caviation was not advanced. The flow distribution through the perforated outer shell should be optimized.

Study on In-vessel ISI for JOYO - Core support plate inspection equipment under sodium -

Ariyoshi, Masahiko; Nakayama, Oukatsu

JNC TN9400 2004-015, 27 Pages, 2004/03

JNC-TN9400-2004-015.pdf:0.8MB

This report describes the feasibility study on the in-vessel inspection technique to be applied for the experimental fast reactor JOYO. Since JOYO is a sodium cooled fast reactor and the core structure components are in high temperature and fast neutron irradiation environment, material strength yields to decrease due to heat fatigue and irradiation induced embrittlement. The present surveillance method of testing structural material mechanical property, monitoring subassembly outlet temperature and sodium leak are not sufficient to detect a small crack. Therefore, the direct inspection method to confirm the core structure integrity needs to be developed considering the recent innovative technology. Since the core structure has a complicated form and components under sodium are not visible, the manipulator technology is essential which has various positioning functions. The core support plate was selected to be an important inspection item, and the method which combines ultrasonic non-destructive inspection technology and manipulator technology was examined. In this study, the core support plate inspection equipment concept was examined on the assumption that the inspection of the core support plate was executed under sodium. The sensor for the nondestructive testing is built into the inspection assembly which is loaded in most outer core area, and can be accessed to a low-pressure plenum sidewall by the manipulator. The linear array sensor which arranges several piezo-electric elements were applied to the sensor for the nondestructive testing, and for scanning the inspection object in non-contact. It was assumed for the concept of the manipulator that the sensor was positioned by moving of the arm and bending the joint. The technology of the Master Slave Manipulator for Laparoscopic Surgery was applied for the manipulator. A detailed examination is necessary for this concept in the future. And, the development of in-vessel inspection equipment is continued and prot

Experimental study on thermal stratification phenomena in sodium-cooled fast reactors; Behavior of thermal stratification interface near dipped plate

Nakayama, Okatsu; Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki

no journal, , 

no abstracts in English