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The use of check valves in the nuclear industry is of great importance from a safety precaution point ofview (McElhaney, 1995). Choosing check valves for these high-pressurized systems comes with agreat challenge. The valves causes what is called check valve slams when closing, leading to a hugepressure wave traveling through the system. To prevent this from happening calculations have to bedone to see what kind of forces are generated during a check valve slam. When the forces are known itis easier designing systems that will endure these slams. A commonly used software in the nuclearindustry is RELAP5 (Reactor Excursion and Leak Analysis Program), its main purpose is to calculatetransients in piping systems. This program can also be used when calculating a check valve slam. Buthow precise is the code compared to the real event? By doing an experiment measuring pressures created by swing check valves during slams, the codewas compared to real data and analyzed to decide what was of importance when modeling for thesetypes of simulations. The RELAP5 code was not initially designed to calculate transients during a check valve slam. This isclearly shown when the code overestimates the pressure waves in the system when using themanufacturer data for the check valve model. Matching the data from the simulations in RELAP5 withthe data recorded from the experiment is not easy. The parameters used for this have no connection tothe specifications for the check valve, which means that transients are hard to estimate withoutexperimental data.

The flow distribution in the primary circuit of the pressurized water reactor was studied with experiments and Computational Fluid Dynamics (CFD) simulations. The main focus was on the flow field and mixing in the downcomer of the pressure vessel: how the different factors like the orientation of operating loops, the total loop flow rate and the asymmetry of the loop flow rates affect the outcome. In addition to the flow field studies the overall applicability of CFD methods for primary circuit thermal-hydraulic analysis was evaluated based on the CFD simulations of the mixing experiments of the ROCOM (Rossendorf Coolant Mixing Model) test facility and the mixing experiments of the Paks NPP. The experimental part of the work in work package 3 included series of steady state mixing experiments with the ROCOM test facility and the publication of results of Paks VVER-440 NPP thermal mixing experiments. The ROCOM test facility models a 4-loop KONVOI type reactor. In the steady-state mixing experiments the velocity field in the downcomer was measured using laser Doppler anemometry and the concentration of the tracer solution fed from one loop was measured at the downcomer and at the core inlet plane. The varied parameters were the number and orientation of the operating loops, the total flow rate and the (asymmetric) flow rate of individual loops. The Paks NPP thermal mixing experiments took place during commissioning tests of replaced steam generator safety valves in 1987-1989. It was assumed that in the reactor vessels of Paks VVER-440 NPP equipped with six loops the mixing of the coolant is not ideal. For the realistic determination of the active core inlet temperature field for the transients and accidents associated with different level temperature asymmetry a set of mixing factors were determined. Based on data from the online core monitoring system and a separate mathematical model the mixing factors for loop flows at the core inlet were determined. In the numerical simulation part of the work package 3 the detailed measurements of ROCOM tests were used for the validation of CFD methods for primary circuit studies. The selected steady state mixing experiments were simulated with CFD codes CFX-4, CFX-5 and FLUENT. The velocity field in the downcomer and the mixing of the scalar were compared between CFD simulations and experiments. The CFD simulations of full scale PWR included the simulation of Paks VVER-440 mixing experiment and the simulation of Loviisa VVER-440 downcomer flow field. In the simulations of Paks experiments the experimental and simulated concentration field at the core inlet were compared and conclusions made concerning the results overall and the VVER-440 specific geometry modelling aspects like how to model the perforated elliptic bottom plate and what is the effect of the cold leg bends to the flow field entering to the downcomer. With Loviisa simulations the qualitative comparison was made against the original commissioning experiments but the emphasis was on the CFD method validation and testing. The overall conclusion concerning the CFD modelling of the flow field and mixing in the PWR primary circuit could be that the current computation capacity and physical models also in commercial codes is beginning to be sufficient for simulations giving reliable and useful results for many real primary circuit applications. However the misuse of CFD methods is easy, and the general as well as the nuclear power specific modelling guidelines should be followed when the CFD simulations are made.

The goal of the work described in this report was the experimental investigation of the mixing of coolant with different quality (temperature, boron concentration) in nuclear reactors on the way from the cold leg through the downcomer and lower plenum to the core inlet in a systematic way. The obtained data were used for the clarification of the mixing mechanisms and form a data basis for the validation of computational fluid dynamics (CFD) codes. For these purposes, experiments on slug mixing have been performed at two test facilities, modelling different reactor types in scale 1:5, the Rossendorf and Vattenfall test facilities. The corresponding accident scenario is the start-up of first main coolant pump (MCP) after formation of a slug of lower borated water during the reflux-condenser mode phase of a small break loss of coolant accident (LOCA). The matrices for the experiments were elaborated on the basis of the key phenomena, being responsible for the coolant mixing during pump start-up. Slug mixing tests have also been performed at the VVER-1000 facility of EDO Gidropress to meet the specifics of this reactor type. The mixing of slugs of water of different quality is also very important for pre-stressed thermal shock (PTS) situations. In emergency core cooling (ECC) situations after a LOCA, cold ECC water is injected into the hot water in the cold leg and downcomer. Due to the large temperature differences, thermal shocks are induced at the reactor pressure vessel (RPV) wall. Temperature distributions near the wall and temperature gradients in time are important to be known for the assessment of thermal stresses. One of the important phenomena in connection with PTS is thermal stratification, a flow condition with a vertical temperature profile in a horizontal pipe. Due to the fluctuating character of the flow, this may cause thermal fatigue in the pipe. Besides of thermal fatigue, a single thermal shock can also be relevant for structural integrity, if it is large enough, especially in the case, that the brittle fracture temperature of the RPV material is reduced due to radiation embrittlement. Therefore, additional to the investigations of slug mixing during re-start of coolant circulation, the mixing of slugs or streams of water with higher density with the ambient fluid in the RPV was investigated. The aim of these investigations was to study the process of turbulent mixing under the influence of buoyancy forces caused by the temperature differences. Heat transfer to the wall and thermal conductivity in the wall material have not been considered. Experiments on density driven mixing were carried out at the Rossendorf and the Fortum PTS facilities.

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