ARIGS is one of the programs on the aerosol retention on the tubes surrounding the breach within the secondary side of the steam generator in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST programs. Experimental activities were focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development.
ARISG-I was a step forward in the modeling of the aerosol retention of the steam generator. According to this analysis the main areas where research is needed are: gas jet behavior across the tube bank; particle resuspension, erosion, and/or bouncing; and particle inertial impaction and turbulent deposition under foreseen conditions.
No description available.
Steam generator reliability and performance are serious concerns in the operation of pressurized water reactors. The aim of the SGTR project was to provide a database of fission product retention in steam generator tube rupture sequences and models, which could be applied to estimate the effectiveness of different accident management strategies in these kind of accidents.
The SGTR project made an important step forward to resolve uncertainties of physical models, especially in the aerosol deposition and mechanical resuspension in turbulent flows. There was one sampling at the injection line for the Optical Particle Counter (OPC) aimed at determining the aerosol size distribution and quantifying the mass concentration at the inlet. Within the vessel atmosphere eight samplings were taken to six filters and two cascade impactors, from which the mass concentration exiting the tube mini-bundle was estimated.
The test mini-bundle is a scaled mock-up of the first stage of the steam generator tube bundle. It consists of a squared arrangement housing inside a total of 117 tubes plus four supporting rods placed in the corners. The mini-bundle allows two possible locations of the broken tube. One place is just at the centre of the structure and the other place is three tubes away from the centre.
Reactor cavity flooding is a cornerstone of severe accident management strategy in Swedish type BWRs. In a hypothetical severe accident with core melting and reactor vessel melt-through, it is assumed that the melt ejected into a deep water pool will fragment, quench and form a porous debris bed coolable by natural circulation. If natural circulation cannot remove decay heat produced by the debris, then dryout, reheating and remelting of the debris bed is expected to occur. Attack of molten core materials on the reactor containment base-mat presents a threat to containment integrity. Amount of the heat which can be removed by natural circulation from the debris bed is contingent, among other factors, upon the properties of the bed as porous media. Debris agglomeration and especially formation of “cake” regions can significantly increase hydraulic resistance for the coolant flow and thus negatively affect coolability of the debris bed. If melt is not completely solidified prior to settlement on top of the debris bed, then agglomeration of the debris and even “cake” formation is quite possible.
The SIMECO (SImulation of MElt COolability) test facility consists of a slice type vessel, which includes a semi-circular section and a vertical section, representing the lower head of the reactor vessel. The size of the facility is scaled to be 1/8 of prototypic PWR type reactors. Fig.1 shows a schematic of the facility and Fig.2 shows the main dimensions of the vessel test section. The diameter and height of the test section are 620 mm and 530 mm, respectively. The width of the test section is 90 mm. The front and back faces of the facility are insulated in order to decrease heat losses. The vessel’s wall, represented by a 23-mm thick brass plate, is cooled by a regulated water loop. On the top of the vessel a heat exchanger with regulated water loops is employed to measure the upward heat transfer. The sideways and downward heat fluxes are measured by employing array of thermocouples at several different angular positions. Practically isothermal boundary conditions are provided at pool boundaries. A cable-type heater 3 mm in diameter and 4 m in length is submerged in the pool and provides internal heating. A heat exchanger mounted on the exit of cooling water, is employed to maintain the cooling capacity of the water. The isothermal bath is designed to provide constant temperature. A circulation pump was mounted in order to establish necessary flow rate. One digital and one analog flowmeter were mounted to measure water flow through the wall of the slice, while one analog flowmeter is used to measure the flow in the upper heat exchanger.
Facility is in operation.
FOREVER program at KTH was concerned with the vessel integrity under the molten corium attack in the reactor lower plenum during a severe accident. Total 9 tests were performed in the FOREVER program, to simulate the behavior of the lower head of a reactor pressure vessel (RPV) under different conditions: French steel/American steel, with/without penetrations, with/without gap cooling.
The facility employs a 1/10th scaled lower head (hemispherical in shape and made of SA533B, American reactor steel) of 400 mm outer diameter and 15 mm wall thickness. A cylindrical shell of 15Mo3 German steel, of 400 mm height and thickness of 15 mm, was welded to hemispherical lower head to make a complete vessel.
Facility is in operation.