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During a severe accident of Pressurized Water Reactor(PWR), the core materials was heated, melt located on the lower head of Reactor Pressure Vessel(RPV). With the temperature rise, the corium will melt through the lower head and discharge into the reactor cavity. Those corium will interact with the concrete and damage the integrity of the containment, so some coolability method should used to quench the corium. In order to investigate the progress of MCCI, a MCCI analysis code, that is MOCO, was developed. The MOCO includes the heat transfer behavior in axial and radial directions from the molten corium to the basemat and sidewall concrete, crust generation and growth, and coolability mechanisms reveal the concrete erosion and gas release, which are important for the interaction process. Cavity ablation depth, melt temperature, and gas release are the key parameters in the interaction research. The physical-chemistry reaction is also involved in MOCO code. In the present paper, the related MCCI experiment data were used to verify the models of the MOCO and the calculation results of MOCO were also compared with other MCCI analysis codes. [ABSTRACT FROM AUTHOR]

Abstract: Local heat transfer distribution of liquid sodium turbulent flow in an annulus has been investigated numerically. The high Reynolds number k-ε model and boundary condition of wall function method were applied. The numerical results are in good agreement with some experimental results in open literature. Heat transfer distributions of liquid sodium and water were compared under the same geometry, flow and heating conditions. Meanwhile, influencing factors, including the turbulence, heating and geometry, were studied. Comparing results of different turbulence conditions, the Nusselt number increases with the Reynolds number increase, and the turbulence was mainly influenced on the entrance of the annulus. By analyzing heat transfer distributions under different inlet temperature and heat flux conditions respectively, it can be concluded that this two factors have little influence on the Nusselt number. The gap size and the curvature were considered as the main geometry factors influencing on heat transfer. When the gap size of the annulus increases from 5 to 15 mm, the Nusselt number will increase. While with the same gap size, as the curvature of the annulus changes, the Nusselt number changes very little. [Copyright &y& Elsevier]

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable. [ABSTRACT FROM AUTHOR]

Experiments have been conducted for single-phase water and two-phase air–water flow in a vertical seven-rod bundle with support plates. Measurements were made for two-phase pressure drops under conditions of total mass velocity from 200 to 850 kg m −2 s −1 , quality from 9 × 10 −4 to 0.27 and pressure from 110 to 230 kPa. Single-phase experimental results show that new developed correlations to obtain frictional factors in the vertical rod bundle and loss coefficients of support plates were in good agreement with experimental data. With respect to two-phase results, the homogeneous flow model dramatically overestimated two-phase friction multipliers in vertical rod bundles, while the new developed correlation presented good predictions. When calculating two-phase pressure drops in support plates, a new defined two-phase multiplier of local pressure drop in support plates was suggested. The two-phase multiplier data of support plates, in which an obvious mass velocity effect was observed, could be fit well in terms of Martinelli parameter and a dimensionless mass velocity. The new correlation developed for two-phase multipliers of support plates was successfully in predicting the experimental data. [ABSTRACT FROM AUTHOR]

In this study, a numerical analysis code (DETAC, Detonation Analysis Code) for hydrogen detonation during the reactor severe accident was developed using Fortran 90 language, and the simulation was performed for the hydrogen detonation. A global-chemistry model was adopted to simulate the chemical reaction. The Euler equations were solved using third-order Runge-Kutta method with fifth-order weighted essentially non-oscillatory scheme handling the convection flux. Afterward, the hydrodynamics solver was verified by comparison of predicted results and exact solutions of four cases of shock tube problems. A hydrogen detonation in a pipe was simulated to verify this code by comparing the results with the classical C-J theory. Furthermore, this code was applied to the hydrogen detonation analysis in the compartment of BWR building. Two cases with different ignition locations were analyzed in this paper and the maximum pressure of these cases were 7.5 MPa and 8.0 MPa, respectively. The pressure and the temperature during detonation were affected by the ignition location. The results indicated that the possibility of reactor building destruction exists if the hydrogen detonation occurs. [ABSTRACT FROM AUTHOR]

• A series of numerical equations and calculational methods are investigated and concluded. • The equations are implemented into the calculation of the natural circulation characteristics. • The experiments are conducted with 3 × 3 rod bundles, and the experimental data is used to validate the code. • The phenomenon of the natural circulation characteristics under motion conditions are analyzed with both the experimental and calculational results. Natural circulation system is very important in the nuclear reactor system, especially for the safety in the operation of nuclear reactor. The natural circulation characteristics are different when the reactor is based in the ocean rather than on land. In this study, a set of numerical models is established on the natural circulation characteristics of fuel rods nuclear power system under motion conditions. The models are implemented into the calculation code compiled by Fortran 90, which includes the mass flowrate model of the natural circulation system, the heat exchanger model, the preheater model, the tube model and the Nitrogen pressurizer model. The code is validated with the experimental investigation and calculation error is within ±12%, which suggests that the code can simulate the natural circulation characteristics under stationary and motion conditions accurately. The analysis includes the natural circulation characteristics under stationary, rolling, heaving and inclining conditions. The influential factors include the rolling period, rolling angular acceleration amplitude, rolling angular amplitude, heaving period, heaving acceleration and inclining angles. The calculation results provide the reactor with an evaluation of reactor natural circulation system characteristics. [ABSTRACT FROM AUTHOR]

Abstract: The once-through steam generator is an essential component of the integrated nuclear power system. A compressible model with fixed boundaries was applied to study the thermo-hydraulic characteristics of an annular tube once-through steam generator. The present model follows the three fundamental conservation principles in the lumped parameters form, and adopts special method for the primary flow distribution in the annular tube. The Gear algorithm is adopted to solve the differential equations. The steady-state and transient results show that the proposed model can successfully simulate the transient processes with large disturbances in the steam generator. Comparisons between the compressible and incompressible flow models indicate that the compressible flow model is necessary to the steam generator transient simulation. [Copyright &y& Elsevier]