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The main purpose of the current study is to numerically investigate, through computational fluid dynamics modeling, a water jet injected vertically downward through a straight circular pipe into a water bath. The study also aims to obtain a better understanding of jet behavior, air entrainment and the dispersion of bubbles in the developing flow region. For these purposes, three dimensional air and water flows were modeled using the volume of fluid technique. The equations in question were formulated using the density and viscosity of a 'gas-liquid mixture', described in terms of the phase volume fraction. Three turbulence models with a high Reynolds number have been considered i. e. the standard k-e model, realizable k-e model, and Reynolds stress model. The predicted flow patterns for the realizable k-e model match well with experimental measurements found in available literature. Nevertheless, some discrepancies regarding velocity relaxation and turbulent momentum distribution in the pool are still observed for both the standard k-e and the Reynolds stress model.

In this article, we present the results of a few energy groups calculations for the NUR reactor fuel depletion analysis up to 45 000 MWd/tU taken as the maximum fuel burn up. The WIMSD-4 cell code has been employed as a calculation tool. In this study, we are interested in actinides such as the uranium and plutonium isotopes, as well as fission products Xe-135, Sm-149, Sm-151, Eu-155, and Gd-157. Calculation results regarding the five energy groups are in a good agreement with those obtained with only two energy groups which can, therefore, be used in all subsequent calculations. Calculation results presented in this article can be used as a microscopic data base for estimating the amount of radioactive sources randomly dispersed in the environment. They can also be used to monitor the fuel assemblies inventory at the core level.

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

There is a common understanding among thermal-hydraulic experts that the system analysis codes have currently reached an acceptable degree of maturity. Reliable application, however, is still limited to the validated domain. There is a growing need for qualified codes in assessing the safety of the existing reactors and for developing advanced reactor systems. Under conditions involving multi-phase flow simulations, the use of classical methods, mainly based upon the one dimensional approach, is not appropriate at all. The use of new computational models, such as the direct numerical simulation, large-eddy simulation or other advanced computational fluid dynamics methods, seems to be more suitable for more complex events. For this purpose, the European Commission financed NURESIM Integrated Project (as a part of the FP6 programme), was adopted to provide the initial step towards a Common European Standard Software Platform for Modeling, recording and recovering computer data for nuclear reactor simulations. Some of the studies carried out at the University of Pisa within the framework of the NURESIM project are presented in this paper. They mainly concern the investigation of two critical phenomena connected with jet instabilities and direct contact condensation that occur during emergency core cooling. Through these examples, the state-of-the-art and the need for model improvements and validation against new experimental data for the sake of getting a better understanding and more accurate predictions are discussed.

RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this purpose, a series of experiments were performed for benchmarking RELAP5 calculations for research reactor systems. As a result of this study, the UMLRR nodalization is shown to be representative of the experimental data reactor behavior.

Prior to core reloading, planned power upgrading, or as a part of required analyses of past events, accurate safety evaluations should be carried out. Generally speaking, the content of a safety report has to be modified whenever a new type or design of fuel is to be used in a reactor core. As the existing plants have well established licensing procedures, including well founded analysis methods, the application of new analysis methods has to be thoroughly evaluated, with specific emphasis on their capability of producing results beneficial to reactor operation. The detailed study presented here was carried out so as to insure that the allowed operational safety limits of the NUR research reactor are not exceeded under any circumstances.

This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to simulate the interactions between the generated steam line pressure wave propagation and the instantaneous core void distribution. An overall comparison shows good agreement between the code calculations and the experimental data. A series of sensitivity analyses were also performed in order to assess the code prediction features, as well as to identify uncertainties related to the adopted thermalhydraulic parameters used for the plant modelisation.