Experimental study of thermal and hydraulic characteristics of VK-300 reactor in solitary uptake tube model
Reactor VK-300 is more powerful then VK-50 and it is interesting for building regional nuclear heart and energy plant. VK-300 is boiling water reactor with integrated concept of equipment, in-vessel steam separation, single-loop steam cycle.
This article is concerned with realization of experimental works of thermohydraulic reactor characterization on the model of single draft tube under pressure 3.4 MPa, different discharge and relative enthalpy on the entry of the model from –0.05 to 0.2. The routine of experiment is: a) creation of circulation of water-steam mixture with target discharge and relative enthalpy through the working unit under pressure 3.3 – 3.4 MPа; b) detection of thermohydraulic parameters of riser and down comer circuits which interesting for practice.
Executed experimental work confirmed the project design which provide the functional ability of reactor and thermohydraulic characteristics of natural circulation loop for reactors calculations with different codes. It is shown that in all range studied of the value of relative enthalpy there are annular-dispersed or droplet-dispersed flow of water-steam mixture in draft tube. It provides the significant separation on the axial separators (the second stage of separation) in the range of moving fall of natural circulation. For the increase of the coefficient in draft tubes in this structure of flow it makes most sense to do the perforation in the top of the draft tube. It will increase the coefficient of separation on the first stage of separation and create more favorable conditions for separation on the second stage.
Measured values of the steam volume fraction in the mixing chamber and the draft tube are in satisfactory agreement with values calculated by the Z.L. Miropol’skij method and code RELAP and can be used for verification accounting code, which used for calculation of thermohydraulic characterizations of VK-300.
It has been shown that it is possible ingress of steam into the ring slit, modeling inter tubular space and its penetration into the nuclear reactor core inlet. The cushion height of steam on this segment depends from its hydraulic resistance. It should be further study of this effect at the operating pressure of units (70 MPa) and various discharge simulating starting and transient modes, for its guaranteed exclusion, as well as the development of emergency response procedures to ensure the safety of the reactor.
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