Measurement of Void Fraction Distribution in Steam-Water Two-Phase Flow in a 44 Bundle at 2 MPa

Journal Article · Wed Jun 15 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:22992027

Liu, Wei; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki ^[1]

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1. Development Group for Thermal-Hydraulics Technology, Nuclear Science and Engineering Center, Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai, Japan, 319-1195 (Japan)

To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam-water two-phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition (Up to 2.8 MPa, 230 deg. C), focusing on the low flow rate condition after the reactor scram. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift-flux models implemented in the two fluids codes, such as TRACE code. We are still in the middle of the data analyses and here we only report a part of the results. For the present bundle rod geometry, the void fraction shows easily gathered at the center area surrounded by 4 rods and the void fraction in the sub - channels between two adjacent rods are low. Under the same steam vapor flow rate condition, void faction increase with the decrease of the water flow rate. Under the same water flow rate, the void fraction increases with the increase of the steam vapor flow rate. In recent future, we will perform more experiments under further low water flow rate condition, including stagnation condition, to confirm this changing tendency. In the present experimental range, no effect of pressure on void fraction was observed. A 3-layer WMS, with a vertical distance between the adjacent layers of 1.3 mm, was used to measure the local instantaneous void fraction in this research. By using the correlation coefficient method, we can determine the time delay between the bubble detections from the upper and the lower layer sensors. Then the speed of the bubble, and its length, can be calculated. We are working on the establishment of the databank for the slug-churn bubbles information. (authors)

• OSTI ID: 22992027
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Journal Issue: 1 Vol. 114; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
DATA ANALYSIS
FLOW RATE
FUEL ELEMENT CLUSTERS
FUKUSHIMA DAIICHI NUCLEAR POWER STATION
REACTOR ACCIDENTS
SCRAM
TEMPERATURE RANGE 0400-1000 K
TWO-PHASE FLOW