Search

The effectiveness of in-vessel retention (IVR) of core melt during severe accident in light water reactor strongly depends on whether the heat removed from boundaries could exceed the heat generated in the molten pool in lower head of reactor pressure vessel (RPV). Natural convection in the oxidic pool plays a key role in determining the heat transfer behavior inside the molten pool. A Molten pool Oxide-layer heat tRaNsfer experiment (MORN) facility was built to investigate the heat transfer characteristic of the molten oxidic pool with the intention to verify the heat transfer correlation of the pool. The test facility consists of a three dimensional hemisphere test section with an inner radius of 0.4 m. The outer boundary of the test section could be cooled by different methods (radiation or convection), so as to study the influence of external cooling condition on the crust formation of oxidic pool. Three kinds of material have been chosen as the simulants. They are water, NaNO3/KNO3(mole ratio 1:4) and a non-eutectic binary mixture of calcium-boron oxide (30 wt%CaO-70 wt%B2O3) respectively. At first period, several water and salt experiments with the melt height of 340 mm were performed and the data were analyzed. Though the normalized temperature distribution of the pool was nearly the same between water and salt experiments, the heat flux across the side wall differed much. The heat flux distribution of the water experiment rose nearly linearly with polar angle till to the melt surface. The heat flux of salt experiment at the low angle was larger than that of water, decreased slightly with angle increasing and then rose quickly till to the melt surface. This is consistent with the LIVE and COPRA experiments with the same boundary condition. The maximum normalized heat flux of the water experiment was approximate 1.6 and that of the salt experiment was about 2.2. The downward heat transfer of MORN is smaller than the correlation derived by COPO, mini-ACOPO and BALI experiments, but is consistent as LIVE, COPRA experiment.

The core make-up tank (CMT) is a crucial device of the passive core cooling system in the Westinghouse AP-series pressurized water reactor nuclear power plants. In the prototype plant, the CMT is a vertical cylindrical vessel with hemispherical top and bottom end sockets. Although the structure of the CMT system is simple, it involves many thermal-hydraulic phenomena, most of which significantly impact the peak temperature of the fuel cladding and the liquid level of the core. Since the CMT not only provides full-pressure injection but also triggers the auto depressurization system by its liquid level signal during small-break loss of coolant accident (SBLOCA), its thermal hydraulic behaviors affect the performance of the passive core cooling system (PXS). Therefore, it is necessary to carry out separate effect tests to verify the feasibility of CMT applied to passive safety and injection systems and to study the operating characteristics of CMT. This article first presents the design of a CMT separate effect test facility and analysis of the operating characteristics of the CMT with the experimental data. The influence of pressure, non-condensable gas, wall condensation, and thermal resistance on the CMT behavior were investigated. The study found that the non-condensable gas would stratify in the CMT, which had negligible effect on the wall condensation, but could greatly reduce the gas-liquid interface condensation.

CAP1400 has been developed as a new passive safety and generation III + reactor based on AP600/1000 technology. Like previous reactors, the small-break loss-of-coolant-accident (SBLOCA) is still one of the most important safety problems that have to be considered before it's widely applications. Although extensive theoretical and experimental results of SBLOCA have been reported in previous studies, the dominant SBLOCA transient mechanisms and the important thermal hydraulic phenomena have not been clearly explained, especially in the passive safety system, due to the complexity of the passive safety nuclear reactor geometry coupled with the transient two-phase fluid interactions. In this study, a new constructed thermal hydraulic integral effect test facility, named as advanced core-cooling mechanism experiment (ACME), which is designed and constructed for supporting CAP1400 safety review, was used to investigate the SBLOCA transient in passive safety plants. 1-inch, 2-inch and 8-inch cold leg breaks and a double-ended direct-vessel-injection (DEDVI) line break were selected as the critical and essential SBLOCA tests. The system pressure, injection flow rate, core level and fluid temperature were considered as the safety related parameters to investigate the system responses to the SBLOCA. In addition, the conditions that affect the SBLOCA chronology were investigated as well. Also, a few important thermal hydraulic phenomena observed in the ACME SBLOCA test were reported, and their effects on the transient were analyzed. The dominant SBLOCA transient mechanisms in the passive safety system were presented. This study not only helps to improve the understanding of SBLOCA mechanisms and important thermal hydraulic phenomena in passive safety systems, but also sheds some lights on SBLOCA analysis, test facility scaling, safety performance evaluation, code simulation and even the design optimization for the passive safety PWRs.

Cold crucible induction melter (CCIM) technology has been widely used in melting of non-metallic refractory materials with a high melting point and a requirement of high purity. This paper discusses the application of the CCIM technology in melting of reactor corium (simulated as zirconium oxide in this experiment). In this experiment, the power supply is set at constant power control mode, with a design frequency of ∼100 kHz and a total output power of ∼ 400 kW. The cold crucible consists of copper tubes and a stainless bottom plate which has a dimension of 220 mm in diameter and 420 mm in height. The crucible is charged with zirconium oxide (ZrO2) powder and surrounded by a set of water-cooling induction coils which is designed shorter than the crucible in order to reduce energy loss in the water-cooling bottom plate. The top of the crucible and the coils are at the same height. A Zirconium ring is used to initiate heating process. ∼ kg ZrO2 powder is used for each experiment. Electrical variables of the induction melter and temperature of the melt are detected and recorded in the entire heating process. For the result, almost 3/4 of the ZrO2 powder is melt and a thin crust (∼ 2–3 mm) is generated between the melt and the crucible. To be specific, the ZrO2 powder above the bottom of the induction coils is melt, but others are not. In the melting process, the maximum surface temperature of the melt has reached 2230°C. Furthermore, for better understanding of the relationship between the melting process and the changing process of the electrical variables, an equivalent circuit of the heating system is established and analyzed. It can be concluded that some electrical variables such as the quality factor (Q) of the induction coils and the practical frequency can reflect melting status. The conclusion is confirmed by the records transmitted by sensors in the melting process. By comparison between the recorded data and the melting status, it is clear that the Q value can be accurately used in indicating the melting status due to its obvious change in the entire melting process. As the melting pool becoming larger, the Q value tends to be smaller. In contrast, the practical frequency becomes higher as the melt becoming larger. Furthermore, both of them are not influenced by the output power. In addition, the output current and the output voltage also can be used to describe the melting status. In specific, a larger melting pool corresponds to a lower output current and a higher output voltage.

In large reactors, the fuel handling including spent fuel storage is solved by deep, lined channels and a spent fuel pool. Since the cores and pressure vessels of heating reactors are largely due to their low volumetric power density and integrated arrangement of primary heat exchangers, in-vessel storage will be a most attractive solution to achieve the economic aim by simplifying a classic refueling scheme. Furthermore, the experiences of long term fuel storage at higher temperatures and reaching very close to the active core will be gained by investigating the effects and data of fuel bundles before discharging. This paper presents an in-vessel storage scheme developed by INET (Institute of Nuclear Energy Technology).

A 5 MWt experimental nuclear heating reactor (NHR) was started at the Institute of Nuclear Energy Technology (INET) in 1989 and has operated successfully for space heating for four winters. Therefore, a commercially demonstrated NHR with output of 200 MWt has been developed by INET. This paper presents the safety characteristics of NHRs. It is proved that the reactor core would always be flooded under water. Some features of the integrated arrangement, hydraulic control rod drives and storage of spent fuel assemblies in a pressure vessel are described.