Modeling of the water level swell during depressurization of the reactor pressure vessel of the boiling water reactor in accidental condition

Abstract

Evaluation of the two-phase water mixture level in the case of the sudden depressurization of the Reactor Pressure Vessel resulting from an accident scenario is an important aspect in the reactor safety analysis. This paper discusses results of simulations of the water dynamics and heat transfer during the process of an abrupt depressurization of a vessel filled up to a certain level with saturated liquid water and with the rest of the vessel occupied by steam under saturation conditions. During the pressure decrease e.g. due to a break in the steam pipeline, the liquid water evaporates abruptly leading to strong transients in the vessel. These transients and the sudden emergence of void in the area occupied by liquid at the beginning, result in the elevation of the two-phase mixture. This work presents several approaches for modelling of the void fraction, the level swell and the collapse level. The first approach was based on the churn turbulent drift-flux correlation and an explicit analytic equation for the averge void fraction as a function of dimendsionless superficial vapor velocity. The second and the third aproaches were based on dimensionless analysis and purely empirical corelations. The models were verified against independent experimental data. The models represent the Reactor Pressure Vessel of the Integral Test Facility Karlstein (INKA) -a dedicated test facility for experimental investigation of KERENA -a new medium size Boiling Water Reactor design of Framatome. The comparison of the simulations results against the reference data shows a good agreement.

Modeling of emergency condenser system response to loss of coolant accident in a BWR III+ generation

“…In order to simulate the decay heat of the reactor core, the vessel is fed with steam by the Benson boiler with maximum power output of 22 MW [3,5,9]. Cross section through the KERENA containment [3] One of the purposes of the test facilities of Framatome is to conduct experiments which contribute to validation of codes and models, which may be used in the future to perform pre-calculations before tests, in some cases to eliminate the necessity of conducting test or to help in the design process of new solutions. The current state-of-art codes used for thermal-hydraulic calculations in nuclear industry have been developed for decades.…”