During a severe accident, contact of the molten corium with the coolant water may cause an energetic steam explosion which is a rapid increase of explosive vaporization by transfer to the water of a significant part of the energy in the corium melt. This steam explosion has been considered as an adverse effect when the water is used to cool the molten corium and could threaten reactor vessel, reactor cavity, containment integrity. In this study, TROI TS-2 and TS-3 experiments as part of the OECD/SERENA-2 project were analyzed with TEXAS-V. Input parameters were based on actual TROI experiment data. In mixing simulations, calculated results were compared to melt front behavior, void fraction in trigger time and other parameters in experiment results. In explosion simulations, corresponding to TROI experiments an external triggering was employed at the moment that melt front reached heights of 0.4 m. Calculated results of peak pressure and impulse at the bottom were compared with TROI experiment results. Melt front behaviors of the melt was different from the experimental results in both TS-2 and TS-3. Void fraction in triggering time in TS-2 was in good agreement with the experiment results and in TS-3 was slightly overestimated. The peak pressure and impulse at bottom were successfully predicted by TEXAS-V. These calculations will allow establishing whether the limitations and differences observed in the simulations of the experiments are important for the reactor case.

A passive injection test was conducted using a core makeup tank (CMT), a safety injection tank (SIT) and an automatic depressurization system (ADS), which consists of a passive safety system (PSS) of the SMART reactor. This paper investigates the thermal-hydraulic interaction between CMT and SIT during sequential injections of coolant from these two tanks to a high-temperature and high-pressure reactor pressure vessel using an integral effect test facility of SMART-ITL (System-Integrated Modular Advanced ReacTor-Integral Test Loop). Both CMT and SIT were connected to the reactor pressure vessel by a pressure balance line (PBL) and injection line (IL). A steady-state condition was maintained for 1,000 seconds before the start of the injection. The major parameters agreed well with the target value. After one of safety injection system line was simulated to be broken, a transient injection test was conducted according to the small-break loss-of-coolant accident (SBLOCA) scenario. Coolant injections from a CMT and SIT were started sequentially by opening quick-opening valves installed on the IL and PBL piping, respectively. Several thermal-hydraulic phenomena such as direct contact condensation, thermal stratification, and coupling effects between the CMT and SIT were locally observed during the SBLOCA scenario. The results show that the adopted passive safety injection system functions well as an emergency core cooling system.

Flow network models of the upper plenum of reactor vessel and the primary inlet plenum of intermediate heat exchanger (IHX) of MONJU were advanced and each model was validated by the measured data obtained in the previous system start-up tests (SSTs). Then the whole plant dynamics in a plant trip transient of MONJU were simulated by incorporating these flow network models. The natural circulation in the primary and secondary heat transport systems were also validated by these test results. From these validations, we concluded that the plant dynamics model incorporated the advanced models in Super-COPD code could simulate the whole plant dynamics in good accuracy both in the transients and natural circulation conditions and that it was applicable to predict the next SSTs.

Advanced light water reactor systems are designed to use passive emergency core cooling systems with horizontal pipes that provide highly subcooled water from water storage tanks or passive heat exchangers to the reactor vessel core under accident conditions. Because passive systems are driven by density gradients, the horizontal pipes often do not flow full and thus have a free surface that is exposed to saturated steam and stratified flow is present. The subcooled water flows into the reactor vessel via the downcomer—which is an annular region separated from the core by a cylindrical, steel, inner liner. Under accident conditions, where the horizontal pipes are only partially full, both saturated water and saturated steam from the downcomer are likely to be present in the pipe. The saturated water is quasi-static and the fraction of the free surface that is saturated water is dictated by the physics of the flow. The saturated steam is usually flowing in the countercurrent direction to the subcooled water. Consequently the flow may be a three-layered system with saturated steam over a static saturated liquid layer over a flowing subcooled layer. The conditions leading to a saturated liquid layer that separates the saturated steam from flowing subcooled water are explored. The variables that influence the formation of the saturated liquid layer and enable the saturated layer to be maintained for a spectrum of conditions, including steam flow in the countercurrent direction to the subcooled water, are derived, and compared to experimental data. Conclusions regarding this type of flow are given. Finally, typical steam flow velocities that may induce wave-bridging, leading to condensation-induced-water hammer, are identified.

The International Standard Problem, ISP 43, was defined by the OECD/NEA and the US NRC for the validation of three-dimensional Computational Fluid Dynamics (CFD) software. The underlying experiment was performed in the 2×4 Loop Facility of the University of Maryland, College Park, U.S.A (UMCP). The test facility is a scaled-down model of the Three Mile Island TMI-2 reactor with detailed reconstruction of the reactor pressure vessel (RPV). The ISP 43 experiments focussed on rapid boron dilution transients in the RPV cold leg and downcomer. The simulations of the ISP 43 were performed with the CFX-TASCflow software. Numerical errors were monitored by comparing results obtained with different higher order discretisation schemes. Uncertainties related to physical modelling, like buoyancy effects and reactor core models, were also investigated. The simulation results show good agreement with data and prove that CFD methods can be usefully applied to this class of nuclear reactor problems.

Within the framework of the nuclear power plant lifetime issue, the assessment of the French 900 MWe (3-loops) series Reactor Pressure Vessel (RPV) integrity was performed. A simplified analysis has shown that one of the most severe loading condition is given by the small break loss of coolant accidents (SBLOCA) due to the pressurized injection of cold water (9°C) into the cold leg and down comer of the RPV. Two main physical phenomena, considered important for the RPV cooling transient, were identified during numerical application obtained with EDF CFD tools. These phenomena are the fluid flow separation and the plume oscillations in the down comer. In order to consolidate these numerical results with the EDF home code, called Code_Saturne , an experimental study has been carried out with the new EDF R&D facility. This transparent experimental model is based on the representation at 1/2 scale of a cold leg and a third of down comer including a thermal shield. The experiments were realized by injecting of salt water flow (density effects) in the cold leg according to a similitude study based on Froude number conservation between experiments and reactor scenarios. Firstly, this paper presents the main qualitative experimental results, based essentially on visualizations of different injections of dyed salt water in the cold leg and in the down comer. The physical phenomena observed showed a qualitative good agreement between visualizations and numerical results. Secondly, this paper presents the first experimental results of the assessment of the fluid flow separation in the experimental model obtained with temperature probes inserted in the down comer. We showed, in the experiments analysis, the fluid flow separation and the jet oscillations were detected. The next step will consist to compare these quantitative experiments with numerical study which will be carry out with Code_Saturne .

The work was aimed at the experimental investigation and numerical simulation of coolant mixing in the downcomer and the lower plenum of pressurized water reactors (PWR). For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas model of a German PWR allowing conductivity measurements by wire mesh sensors and velocity measurements by LDA technique. The CFD calculations were carried out with the CFD-code CFX-4. For the design of the facility, calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. Flow measurements and the corresponding CFD calculations in the ROCOM downcomer under steady state conditions showed a Re number independency at nominal flow rates. The flow field is dominated by recirculation areas below the inlet nozzles. Transient flow measurements with high performance LDA-technique showed in agreement with CFX-4 results, that in the case of the start up of a pump after a laminar stage large vortices dominate the flow. In the case of stationary mixing, the maximum value of the averaged mixing scalar at the core inlet was found in the sector below the inlet nozzle, where the tracer was injected. At the start-up case of one pump due to a strong impulse driven flow at the inlet nozzle the horizontal part of the flow dominates in the downcomer. The injection is distributed into two main jets, the maximum of the tracer concentration at the core inlet appears at the opposite part of the loop where the tracer was injected. For turbulent flows the CFD-Code CFX-4 was validated and can be used in reactor safety analysis. Due to the good agreement between measured results and the corresponding CFD-calculation efficient modules for the coupling of thermal hydraulic computer codes with three-dimensional neutron-kinetic models using the results of this work can be developed. A better description of the mixing processes inside the RPV is the basis of a more realistic safety assessment.

Within the Reactor Pressure Vessel (RPV) plant life time project, Electricite´ de France (EDF) must carry out studies showing the RPV integrity during severe loading conditions as the Small Break Loss Of Coolant Accidents (SBLOCA) scenarios. A recent thermal-hydraulic SBCLOCA scenario study, concerning a certain RPV geometry, showed two-phase-flow conditions appears in the cold leg. The single phase flow CFD tools, usually used in these cases, are unsuitable to carry out such studies. That is the reason why EDF is developing a new CFD code (called Neptune _ Code ) implementing two-phase flow models. In order to improve the physical phenomena knowledge and the existent experimental data for Neptune _ Code development , EDF carried out an experimental study concerning the stratified flow (salt water) and free surface (water air) behavior during a representative ECC scenario with partial uncovered cold leg. This paper focuses mainly on the experimental method and the measurements obtained in three tests showing the effects of the uncovered cold leg level on the stratified flow and the free surface behavior.

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.