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| Content Provider | The American Society of Mechanical Engineers (ASME) Digital Collection |
|---|---|
| Author | Morozov, A. V. Remizov, O. V. Kalyakin, D. S. |
| Copyright Year | 2012 |
| Abstract | In new Russian NPP with VVER reactor in the event of LOCA, provision is made for the use of passive heat removal system for necessary core cooling. In the case of leakage in the primary circuit this system assures the transition of steam generators to operation in the mode of condensation of the primary circuit steam coming to steam generator piping from the reactor. As a result, the condensate from the steam generators arrives to the core providing its additional cooling. The steam generator off-design condensation mode has the following features: undeveloped nucleate boiling on the horizontal tubes heated by condensing steam; natural circulation processes in the both steam generator circuits; low heat fluxes and temperature differences. The experimental study of undeveloped nucleate boiling on the single horizontal tube heated by condensing steam has been carried out in the Institute for Physics and Power Engineering. The experiments have been carried out on the GROT test facility. The heart of the test facility is the single VVER steam generator tube (length l = 10.2 m, outer diameter D = 16 mm, wall thickness d = 1.5 mm). The tube is fabricated of the original stainless steel 08Cr18Ni10Ti. The length and geometry of the test tube corresponded to that of real steam generator. The test facility was equipped with thermocouples enabling the temperatures of primary and secondary facility circuits to be controlled. The experiments were carried out at three heating steam pressures Ps1: 0.21, 0.35, 0.55 MPa. The main task of the research was to study the pressure effect on the process of undeveloped nucleate boiling on the single horizontal tube. On the base of the results of these experiments the empirical correlations for prediction of heat transfer coefficient and heat flux were obtained. The generalizing empirical correlations obtained can be used for the substantiation of work heat-exchanging equipment of NPP with VVER reactor in the condensation mode, and also can be applied to the verification of computer codes. |
| Sponsorship | Nuclear Engineering Division Power Division |
| Starting Page | 573 |
| Ending Page | 578 |
| Page Count | 6 |
| File Format | |
| ISBN | 9780791844977 |
| DOI | 10.1115/ICONE20-POWER2012-54809 |
| Volume Number | Volume 3: Thermal-Hydraulics; Turbines, Generators, and Auxiliaries |
| Conference Proceedings | 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference |
| Language | English |
| Publisher Date | 2012-07-30 |
| Publisher Place | Anaheim, California, USA |
| Access Restriction | Subscribed |
| Subject Keyword | Computers Test facilities Temperature Circuits Thermocouples Condensation Heat flux Design Condensed matter Nucleate boiling Leakage Cooling Boilers Wall thickness Pressure Physics Geometry Heat Energy engineering Heating Stainless steel Steam Flux (metallurgy) Nuclear power stations Heat transfer coefficients Pipes |
| Content Type | Text |
| Resource Type | Article |
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