COPRA: A large scale experiment on natural convection heat transfer in corium pools with internal heating. (January 2016)

• Record Type: Journal Article
• Authors: Zhang, Luteng; Zhang, Yapei; Zhao, Bo; Ma, Weimin; Zhou, Yukun; Su, G.H.; Tian, Wenxi; Qiu, Suizheng
• Is Part Of: Progress in nuclear energy. Volume 86(2016:Jan.)
• Page Start: 132
• Page End: 140
• Publication Date: 2016-01
• Subjects: Severe accident -- Natural convection -- Corium pool -- IVR -- COPRA; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 333.7924
• DOI: 10.1016/j.pnucene.2015.10.006
• Languages: English
• ISSNs: 0149-1970

Abstract

Abstract: During a severe accident in light water reactors, the core may melt and relocate into the lower plenum of the reactor pressure vessel. The decay heat will threaten the structural and thermal integrity of the reactor vessel if there is no effective cooling mechanism. Natural convection plays an important role in determining the thermal-hydraulic behaviors inside the debris pool, which is directly relevant to the problem of retention of molten corium inside the lower plenum. The COPRA (COrium Pool Research Apparatus) experiments were performed to study the natural convection heat transfer behavior in an internally heated melt pool with high Rayleigh numbers. The COPRA test facility is a two-dimensional 1/4 circular slice vessel with an inner radius of 2.2 m to simulate the lower plenum of reactor vessel for the Chinese advanced PWR in a full scale. The inner width of the slice is 20 cm and the curved vessel wall has a thickness of 30 mm. 20 electrical heating rods, each with a diameter of 16 mm but different lengths according to their locations, are uniformly distributed in the vessel to simulate homogenous internal decay heat. They can provide a maximum of 30 kW power to the melt pool. The outside of the curved wall is enclosed with a regulated external cooling path to keep the boundary temperature nearly isothermal. The top surface of the pool can be maintained insulated with an adiabatic lid. 79 thermocouples are installed in the melt pool to measure the melt pool temperature field and 48 in the curved wall to obtain local heat flux distribution along the curved wall. In the first series of experiments, water was employed as the simulant material. Due to the full scale geometry, the Rayleigh number within the pool could reach up to 10 16, matching that in the prototypical situation for PWR. The heat transfer characteristic with volumetrically internal heat was investigated by using the full scale facility COPRA. Relations of pool temperature and heat flux distribution, as well as Nudn -Ra′ were developed. The results have been compared with the results and correlations from other experiments. Highlights: The large scale COPRA experiments were performed to study the heat transfer characteristics in corium pools. The Rayleigh numbers could reach to 10 16 to cover the prototypic range of current interest of in-vessel retention. Relations of pool temperature and heat flux distribution, as well as Nudn -Ra' were developed.