| 1. |
- Giri, Asis, et al.
(författare)
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Analysis of bubble dynamics in explosive boiling of droplet with fine fragmentation
- 2005
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Ingår i: Experimental Thermal and Fluid Science. - : Elsevier BV. - 0894-1777 .- 1879-2286. ; 29:3, s. 295-303
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Tidskriftsartikel (refereegranskat)abstract
- If a high temperature liquid comes into contact with cold and volatile liquid, rapid (or explosive) evaporation may occur spontaneously or triggered by the impact of a pressure wave. This event generating a shock wave is called a steam explosion. It involves many multiphase flow and heat transfer phenomena. One of the more important phenomena in a steam explosion is the fine fragmentation of the hot liquid, which determines the explosive heat transfer from the hot liquid to the cold liquid and the vaporisation rate of the cold liquid. When a small hot single drop (similar to1 mm) interacts with the coolant, a vapour bubble is formed around the drop. It was observed experimentally that these vapour bubbles grow and collapse. During this process, the small hot droplet fragments and generates finer particles. To understand the fine fragmentation process during a steam explosion, in this study, this phenomenon was examined by using non-linear stability analysis of vapour bubble dynamics based on a concept developed by Inoue et al. [Chem. Eng. Commun. 118 (1992) 189]. From the analysis, it was observed that higher spherical modes were very much unstable during collapse process, which decided the size of the fragmented particles. Vapour shell between molten metal and coolant was considered unstable if the amplitude of one of the spherical modes was greater than vapour shell thickness. In addition, the mass of fragmented particles during each cycle of vapour bubble dynamics was predicted from the analysis. The calculated results were found to be in reasonable agreement with the previously reported [J. Non-Equil. Thermodyn. 13 (1988) 27] experimental results.
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| 2. |
- Sehgal, Balraj, et al.
(författare)
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Assessment of reactor vessel integrity (ARVI)
- 2005
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 235:04-feb, s. 213-232
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Tidskriftsartikel (refereegranskat)abstract
- The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.
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| 3. |
- Sehgal, Balraj, et al.
(författare)
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Experiments on in-vessel melt coolability in the EC-FOREVER program
- 2004
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Ingår i: Proceeding of the 2004 international congress on advances in nuclear power plants (ICAPP'04). - 0894486802 ; , s. 917-927
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Konferensbidrag (refereegranskat)abstract
- This paper reports the results from the experiments conducted on the coolability of corium melt during a severe accident scenario when the bottom head is full of the core melt, undergoing natural circulation. These experiments are part of the EC-FOREVER Program in which vessel failure experiments have also been performed. The experiments are performed in a 1/10. scale vessel ({approx_equal} 400 mm dia. and 15 mm wall thickness) and the oxidic melt employed is the mixture CaO + B{sub 2}O{sub 3} at {approx_equal} 1400 K, representing the corium melt mixture of UO{sub 2} + ZrO{sub 2}. The experiments employed an initial phase, during which uniform volumetric heating of the melt was provided and the vessel was pressurized to {approx_equal} 25 bar, for several hours, to generate maximum creep deformation of {approx_equal} 5%; in order to provide the conditions for the formation of a gap between the melt-pool crust and the bottom head wall. After this phase, the vessel was flooded with water. Data was obtained on the vessel and the melt pool temperatures in one of the EC-FOREVER experiments reported here. In the second experiment, additional data was obtained on the steam flow rate and the heat transfer to the water, at the upper face of the melt pool, as a function of time. It was found that the gap cooling mechanism was not effective in reducing the vessel wall temperatures after water flooding. Post test examinations revealed that the water ingression extended to the depth of only {approx_equal} 60 mm in the melt pool.
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