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In the scaled-down test facilities that simulate the accident transient process of the prototype nuclear power plant, the stored energy release in the metal structures has an important influence on the accuracy and effectiveness of the experimental data. Three methods of stored energy analysis are developed, and the mechanism behind stored energy distortion in the test facilities is revealed. Moreover, the application of stored energy analysis is demonstrated for the ACME test facility newly built in China. The results show that the similarity requirements of three methods analyzing the stored energy release decrease gradually. The physical mechanism of stored energy release process can be characterized by the dimensionless numbers including Stanton number, Fourier number and Biot number. Under the premise of satisfying the overall similarity of natural circulation, the stored energy release process in the scale-down test facilities cannot maintain exact similarity. The results of the application of stored energy analysis illustrate that both the transient release process and integral total stored energy of the reactor pressure vessel wall of CAP1400 power plant can be well reproduced in the ACME test facility. [ABSTRACT FROM AUTHOR]

AP1000 nuclear plant is an advanced pressurized reactor developed by Westinghouse Electric Company (WEC), with several passive systems applied to the plant to increase its safety. One of the main passive systems is the passive containment cooling system (PCCS), which aims to ensure the integrity of the containment and remove decay heat from the containment during postulated accidents. Film evaporation outside the steel vessel wall plays an important role for the heat removal process in the PCCS. In this study, the Eulerian Wall Film model in ANSYS FLUENT is applied to study falling film evaporation and natural convection outside the PCCS of AP1000. The thermal–hydraulic phenomena of both liquid film and gas mixture in the outside air flow path of the PCCS are analyzed. The Eulerian Wall Film model shows promising performance in modeling falling film behavior and mass transfer between liquid film and gas phase. [ABSTRACT FROM AUTHOR]

• Obtain the general rules and methods of RELAP5/MOD 3.4 to simulate CMT. • The RELAP5/MOD 3.4 can accurately simulate most of the CMT operating conditions. However, large uncertainty is expected in the simulation of delayed time. • The CMT delayed drainage experiment cannot verify the accuracy of the RELAP5/MOD 3.4 direct contact condensation model. The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during accident and has been widely used in the field of nuclear thermal-hydraulics and safety analysis. The core make-up tank (CMT) is a crucial device of the passive core cooling system in the AP series pressurized water reactor nuclear power plants. Since the CMT not only provides full-pressure injection but also triggers auto depressurization system by its level signal during small-break loss of coolant accident. Therefore, it is necessary to carry out CMT separate effect tests to verify the feasibility and accuracy of RELAP5/MOD 3.4 to simulate CMT. The previous studies found that the lower version of the RELAP5 program has problems in simulating CMT, and the applicability of the latest RELAP5/MOD3.4 has not been evaluated. In this paper, a CMT separate effect test facility were presented and then RELAP5/MOD 3.4 was used to establish a CMT test bench model to evaluate the applicability, and obtain the general rules and methods of RELAP5/MOD 3.4 to simulate CMT. It's found that RELAP5/MOD 3.4 can accurately simulate most of the CMT operating conditions. However, large uncertainty is expected in the simulation of delayed time and the flowrate of delayed drainage stage is not affected by the heat transfer coefficient of steam direct contact condensation. The simulation of delay time depends on experimental data and the CMT delayed drainage experiment cannot verify the accuracy of the RELAP5/MOD 3.4 direct contact condensation model. [ABSTRACT FROM AUTHOR]

Since oxide and metal materials are not miscible and Uranium extraction, the molten reactor core could turn into a multi-layer configuration molten pool in the Reactor Pressure Vessel (RPV). Three-layeR cOupled heat tranSfer molten pool Experiment (TROSE) is established with a three-dimension hemispheric test section. Two preliminary tests are carried out, and these tests use water to simulate both the top and middle layers and Cerrobend alloy to simulate the bottom layer. TROSE results demonstrate that upward heat transfer direction is found at the interface of the middle and bottom layers, and the power split ratio of the middle layer is about 1.0. The top and middle layers have an increasing trend of sideward heat flux; however, the bottom layer shows a different trend due to crust formation. Generally, the calculated results of Churchill-Chu, BALI, and COPRA correlations are close to the experimental sideward heat flux of the different layers in the multi-layer configuration. This three-dimension study provides additional experiment data for the three-layer configuration in IVR strategy for a new Chinese passive nuclear power plant. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The fourth stage of Automatic Depressurization System(ADS4) is a crucial part in the passive core cooling system. During LOCA transient or long-term cooling period, entrainment/offtake phenomenon will take place at the ADS4 branch. This phenomenon may affect the depressurization process or boric acid balancing. The RELAP5 mod3.3 is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. In this case, accurately evaluating the entrainment/offtake model in RELAP5 is quite important for safety analysis of passive nuclear power plant. A F ull sc A le T est facility for liquid E ntrainment ( FATE ) is designed and fabricated to give a better understanding of entrainment's scaling effect and quantify the uncertainty of existing models. The geometry of upper plenum, hot leg and ADS4 pipe are all the same with CAP1400. Test data of both onset of liquid entrainment (OLE) and entrainment rate for different gas and liquid flow rate are achieved. Calculation results are compared with the experiment data and existing entrainment models. The differences are discussed and analyzed. [ABSTRACT FROM AUTHOR]