RALOC4 code validation against measured data at Ignalina NPP during single Main Safety Valve opening

Although RALOC4 code is validated against many experiments with regard to Western Nuclear Power Plants (NPPs) the code validation problem for the Accident Localization System (ALS) of Ignalina NPP modeling is of special importance because the condensing pools at NPP with RBMK-1500 differ from the pressure suppression systems installed in NPPs with German BWR. The response of Ignalina NPP ALS to the unintentional opening of single Main Safety Valve, which occurred in 1998, is analyzed by employing code RALOC4. The results of post-event calculations compared with the measured data available after the event. The performed analysis showed that RALOC4 code could be applied for the simulation of Ignalina NPP ALS. Nevertheless, the spray modeling in RALOC4 should be improved allowing the simulation of sprays in NON_EQUILIBRIUM zone model and to consider the diameter of water droplet diameter and height of droplet fall.     

COCOSYS code validation against measured data during unintended single main safety valve opening at Ignalina NPP

The analysis of an unintended main safety valve opening at Ignalina NPP was performed with COCOSYS code in order to assess its capability in simulation of the transient processes that occur inside Accident Localisation System of Ignalina NPP. COCOSYS has several user-selected options, e.g. zone model (EQUIL._MOD, NONEQUILIB), water flow model (BAL_DRAIN, DRAIN_BOT), etc for nodalisation development. The influence of a zone model selection, a water overflow model selection and efficiency of heat exchanger in Condenser Tray Cooling System was investigated and presented in the paper. The performed analysis supported introduction of new water overflow model in COCOSYS code and showed that COCOSYS code can be applied for the analysis of Accident Localisation System of Ignalina NPP.     

Thermal–hydraulic evaluation of the RBMK-1500 accident confinement system using CONTAIN 11AF

The response of the RBMK Accident Confinement System to a large break LOCA, medium break LOCA and small break LOCA is analyzed using the CONTAIN 11AF code. The effect of Condenser Tray Cooling System failure is investigated for the large break LOCA case. The analysis employs a best estimate mass/energy source and considers both short and long-term responses of the Accident Confinement System. Parametric studies are performed to evaluate the effects of water deposition on the short-term pressure peak and of by-pass leakage on long-term pressure increases.     

Strength evaluation of a steam distribution device in the Ignalina NPP accident localisation system

The Ignalina NPP has a pressure suppression type of confinement, which is referred to as the accident localization system (ALS). The ALS prevents the release of the radioactive material from the NPP to the environment during a loss-of-coolant accident (LOCA). Ten water pools are located in the two ALS towers (five pools in each tower), which separate the dry well from the wet well. These water pools condense the accident-generated steam and prevent high overpressures in the compartments.The steam distribution device (SDD), with the vertical vent pipes (nozzles) that are inserted under the water of the condensing pools, connects the dry well and the wet well. In case of an accident, these components must be capable of withstanding the dynamic loads generated by a LOCA for successful pressure suppression function.This paper presents the transient analysis of the SDD and their connections to the vertical steam corridors following a LOCA. A thermo-hydraulic analysis of the SDD was performed using the state-of-the-art COCOSYS code to determine pressure and temperature histories resulting from a LOCA. The finite element code NEPTUNE was used to evaluate the structural integrity of the SDD and its supporting reinforced concrete wall. Results show that, although portions of the SDD undergo plastic response and the outside surface of the vertical steam corridor reinforced concrete wall cracks, the structural integrity of the SDD and wall are maintained during a LOCA.     

Experimental Investigation of Heat Transfer and Flow Mixing in Pebble Beds

The purpose of this paper is to investigate the thermal hydraulic characteristics of two types of test sections: pebble beds placed between cylinders and thin annular pebble beds (i.e., spheres dumped in thin annular slots). The investigations were performed for axial air flow. The experimental results of heat transfer from the spheres, as well as the results of heat transfer from a cylinder for both types of test sections, are presented. In addition, the results of flow-mixing investigations in pebble beds between axially streamlined cylinders are described. From these results, correlations were developed that can be used to predict the heat transfer characteristics.

The results of the experimental investigation of pebble beds between cylinders demonstrated that the randomly arranged pebble bed is preferable to the regular rhombic structure from the points of view of design simplicity, heat transfer from the cylinder, drag coefficient, and flow mixing. The results of the experiments in the case of thin annular pebble beds demonstrated that the maximum heat transfer from the spheres is at the relative width of the annular slot K, equal to 1.07 and 1.75 of the sphere's diameter     

Reactor cavity and ALS thermal-hydraulic evaluation in the case of fuel channels ruptures at Ignalina NPP

This paper presents the work analysis of the thermal-hydraulic parameters behavior in the RBMK-1500 reactor cavity (RC) and other connected volumes in the case of fuel channels ruptures. The analysis is performed with CONTAIN code using the models of accident localization system (ALS) and reactor cavity venting system (RCVS). The RCVS capacity is assessed and expressed as a number of ruptured fuel channels at which the integrity of RC is maintained. The uncertainty analysis of pressure behavior in RC during multiple fuel channel rupture was performed. The initial and boundary conditions and the code models were selected and their influence on the results is estimated.Calculation of coolant mass and energy release to the reactor cavity in case of fuel channels rupture performed using the main circulation circuit model of Ignalina NPP, which was developed by employing state-of-the-art code RELAP5/MOD3.2 [Fletcher et al., RELAP5/MOD3 code manual user’s guidelines, Idaho National Engineering Lab., NUREG/CR-5535 (1992)]. These results were applied further as the initial data for the analysis of the thermal-hydraulic parameters behavior in the affected compartments employing CONTAIN code.