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Digital archive of theses discussed at the University of Pisa


Thesis etd-08032020-210207

Thesis type
Tesi di dottorato di ricerca
Thesis title
GOTHIC Code Improvement for Phenomena Involving Post Critical Heat Flux Conditions for In-Vessel Retention
Academic discipline
Course of study
tutor Prof. Ambrosini, Walter
supervisore Prof. Forgione, Nicola
supervisore Dott. Petruzzi, Alessandro
  • CHF
  • critical heat flux
Graduation session start date
GOTHIC is an integrated, general purpose thermal hydraulic (TH) software package for design, licensing, safety and operating analysis of Nuclear Power Plant containments, confinement buildings and system components. It bridges the gap between the lumped parameter codes frequently used for containment analysis (like the MELCOR, MAAP, COCOSYS and ASTEC codes) and Computational Fluid Dynamics (CFD) codes.
Within a single nodalization model, GOTHIC can include regions treated in both conventional lumped parameter mode and regions with three-dimensional flows in complex geometries. Although it does not include the capability to model the details of the boundary layers as in most CFD codes, through the use of standard wall functions for heat and momentum transfer, it can give good estimates of the three dimensional (3D) flows and distribution with significantly lower computational cost than typical CFD codes. Additionally, it includes the capability to model multiphase flows situations including dropwise and wall condensation, pool surfaces and sprays without special user supplied models. The 3D capabilities of GOTHIC in simulating basic flows, and in detail, hydrogen flows for containment analysis have been investigated extensively, simulating tests in facilities like PANDA, CSTF, BFMC or CVTR.
The heat transfer correlations built into GOTHIC cover the portion of the boiling curve which spans single phase heat transfer to pre-CHF (critical heat flux) heat transfer. The boiling curve is truncated to exclude post-CHF heat transfer as it has not been adequately verified and was considered by the developers to have little application in general containment analysis. As such, one area that the code is not currently qualified is for In-Vessel Corium Retention analysis, where water enters in contact with the high temperature pressure vessel walls, requiring the modelling of CHF heat transfer. Depending on the heat transfer flux, different amounts of steam may be produced as water enters in contact with the high temperature RPV surface; this will have a direct impact on containment pressure as well as on the integrity of the vessel wall.
The performed doctoral research had the goal of creating an external function/subroutine that resolves this GOTHIC code limitation, enabling it to account for CHF phenomena. Being able to model CHF would represent a significant improvement in simulating the external Pressure Vessel Cooling of different reactor types, as well as other types of severe accidents. With proper heat transfer models and qualification, the modelling capabilities of the GOTHIC code can be expanded to address additional phenomena than envisaged to date.
The developed subroutine and its implementation were made without having access to the GOTHIC source code and was based on the 2006 CHF Look-up Tables by Groeneveld, as one of the most widely used methods for the prediction of CHF. A modifier coefficient was applied to the Look-up Table in order account for different surface angles. Based on experimental data from the ULPU facility, installed at the University of California, Santa Barbara in USA, the subroutine was qualified for CHF situations occurring due to In-Vessel Retention (IVR) accident management procedures using data from 9 different full-scale experiments.
In order to better analyze the performance of the CHF subroutine, the model was applied to the real containment model of Atucha-I, for two severe accident scenarios (LBLOCA and SBO). Compared to a previous analysis of the containment without the CHF model, which used very conservative boundary conditions to model the vapor generation due to the postulated IVR procedure, current results show that the postulated procedure of IVR would be successful with a much lower risk to the Containment integrity due to its lower pressurization (notably lower pressure than the initial analysis, while successfully transferring the imposed decay heat).