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[en] The supercritical water reactor is one of the designs selected for further evaluation by the Gen-IV International Forum. In the framework of the EU FP6 project HPLWR-2 (High Performance Light Water Reactor - Phase 2 ) a concept of the supercritical water reactor is developed. For the safety assessment of new HPLWR concepts, the capabilities of thermal-hydraulic codes to cope with water at supercritical conditions and especially with the transition from supercritical to sub-critical conditions are of crucial importance. The CATHARE 2 and the RELAP5 codes are foreseen to be employed in the HPLWR-2 project, in order to perform safety analyses for the HPLWR design. Therefore, tests have been carried out to assess the capabilities of the selected system codes. (orig.)
[en] The characteristics of flashing-induced instabilities, which are of importance during the startup phase of natural-circulation boiling water reactors, are studied. Experiments at typical startup conditions (low power and low pressure) are carried out on a steam/water natural-circulation loop. The flashing and the mechanism of flashing-induced instability are analyzed. The effect of system pressure and steam volume in the steam dome is investigated as well.The instability region is found as soon as the operational boundary between single-phase and two-phase operation is crossed. Increasing pressure has a stabilizing effect, reducing the operational region in which instabilities occur. Nonequilibrium between phases and enthalpy transport are found to play an important role in the instability process. In contrast with results reported in the literature, instabilities can occur independently of the position of the flashing boundary in the adiabatic section of the loop. The period of the oscillation is found to be about twice the fluid transit time in the system
[en] The capability of the best-estimate thermal-hydraulic code TRACE to correctly reproduce the heat transfer in the core of a Pressurized Water Reactor during the reflood phase of a postulated large break loss of coolant accident was investigated. The code assessment was performed on the basis of reflood experiments carried out at the ACHILLES and NEPTUN experimental facilities. A total of 78 ACHILLES tests and 2 NEPTUN tests were simulated. The typical deviation from the ACHILLES experimental results with respect to the average maximum surface rod temperature and the quenching time was approximately 20degC and 25 s respectively. The rod temperature time history is captured well for the NEPTUN experiments. However, an over-estimation of the quenching time by approximately 10 seconds is observed in the simulations. In general, TRACE is capable of calculating the reflood process with reasonable accuracy. The differences between simulations and experimental data are observed mainly for the upper part of test sections. These differences may be caused by a) lack of a spacer grid model in TRACE; b) insufficient simulating capability for the top quench behavior; c) vessel shroud heat transfer and 3D effects. The analysis points out to the need for more uniform experimental data to exclude significant three-dimensional effects and the influence of the vessel structure surrounding the test section. (author)
[en] Highlights: • Very thin absorber modeling using 2-D step MOC for fuel assemblies. • IFBA requires ray spacing of 50 μm, 32 azimuthal angles to meet 100 pcm criterion. • Homogenization of radial and azimuthal CRUD distributions is investigated. • Up to 300 μm of coolant may be used to homogenize CRUD layer. • Effects of azimuthal homogenization of CRUD is strongly dependent on boron-10 concentration. - Abstract: The method of characteristics (MOC) neutron transport modeling requirements of very thin CRUD and burnable absorber layers on pressurized water reactor (PWR) fuel rods are investigated. Ray tracing parameters, including spacing size and number of angles, and mesh refinement studies are performed for two-dimensional assembly simulations using the DeCART code. It is found that the presence of a 10 μm thick Integral Fuel Burnable Absorber (IFBA) layer within the lattice model requires a ray spacing that is approximately five times smaller than modeling a lattice without IFBA. The modeling of a CRUD layer necessitates less stringent requirements, due to the fact that the boron-10 concentration in a CRUD layer is at least one order of magnitude lower than what is typically found in IFBA layers. Three dimensional CRUD distributions and compositions computed by the CRUD chemistry code MAMBA are extracted and utilized. The reactivity effects of radial and azimuthal mesh refinement, including homogenization strategies of the CRUD layer with the coolant, are investigated. Additionally, the reactivity effect of including the CRUD structure porosity is discussed. A fresh fuel lattice with 1300 ppm is also compared with a reduced boron concentration of 0 ppm to assess the reactivity response of the system to changes in the reactor environment.
[en] During recent years, the best estimate system code TRACE has evolved as the recommended and supported tool by the United States Nuclear Regulatory Commission (USNRC) to simulate Light Water Reactor (LWR) transients. The improvements of version 5.0 are palpable as the code has been able to simulate accurately different scenarios in test facilities and commercial Nuclear Power Plants (NPPs). However, many users still prefer the use of the mature RELAP5 system code as the validation of TRACE is still under way. One of the main areas of interest is the simulation of small break loss-of-coolant-accident (SBLOCA) with boron dilution transients. From 2000 to 2007 different tests have been carried out in the PKL test facility with the purpose to provide valuable experimental data in boron dilution transients. Several organizations have simulated these tests with satisfactory results using different codes. In order to further validate the TRACE code, a well tested RELAP5 model of the PKL test facility developed at the Technical University of Catalonia (UPC) has been translated into TRACE. The results obtained for the PKL tests F1.1 and F1.2, both related to boron dilution issues, are compared. TRACE provided results in very good agreement with both the experiment and the RELAP5 simulation laying out a high matureness of the code. (author)
[en] The prediction of thermal fatigue caused by mixing of fluid streams with different temperature needs fluid-dynamics simulations for the correct estimation of the temperature fluctuations in the wall which is exposed to the temperature patterns generated within this fluid. Currently, most of the efforts in this field are focused on the application of large eddy simulations (LES), which have the disadvantage of high CPU-time requirements and are therefore limited to relatively small geometries. In this paper an attempt to predict temperature fluctuations by means of steady-state RANS simulations is presented. At this aim, the Reynolds stress equations are solved together with a transport equation for the temperature fluctuations. The great advantage of this approach lies in the considerably lower computational requirements if compared to LES and unsteady RANS, therefore allowing the application to more complex and larger geometries, such as the upper and lower head of a reactor pressure vessel. The model is validated against a T-junction experiment performed at the Paul Scherrer Institute, where the mixing patterns between water streams is measured by means of advanced instrumentation with high temporal and spatial resolution, providing transport scalar fluctuations for the validation of the theoretical model. Future efforts will be dedicated to the prediction of the fluctuations frequency, that combined with the fluctuations intensities, provide the boundary conditions for the analyses of thermal stresses in the structures. (author)
[en] An assessment of void-fraction correlations and drift-flux models applied to stationary and transient flashing flows in a vertical pipe has been performed. Experiments have been carried out on a steam/water loop that can be operated both in forced- and natural-circulation conditions to provide data for the assessment. The GE-Ramp and Dix models are found to give very good predictions both for forced- and natural-circulation flow conditions, in the whole range of measured void fractions.Advanced instrumentation, namely, wire-mesh sensors, has been used to obtain a detailed picture of the void-fraction development in the system. On the basis of experimental data, a three-dimensional visualization of the transient flow pattern during flashing was achieved. A transition of the flow pattern between bubbly and slug/churn regimes was found
[en] Highlights: → The spurious opening of 8 relief valves of the ADS system in a BWR/6 has been simulated. → The valves opening results in a fast depressurization and significant loads on the RPV internals. → This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the available plant data. - Abstract: The paper presents the results of a post-event analysis of a spurious opening of 8 relief valves of the automatic depressurization system (ADS) occurred in a BWR/6. The opening of the relief valves results in a fast depressurization (pressure blow down) of the primary system which might lead to significant dynamic loads on the RPV and associated internals. In addition, the RPV level swelling caused by the fast depressurization might lead to undesired water carry-over into the steam line and through the safety relief valves (SRVs). Therefore, the transient needs to be characterized in terms of evolution of pressure, temperature and fluid distribution in the system. This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the plant data.
[en] A variety of nuclear system transients can lead to rapid and large local pressure changes that propagate along the hydraulic system at the speed of sound, both in single phase and in two-phase fluids. Because of the relevance for safety issues, nuclear system codes like TRACE need to be assessed with respect to their capabilities to predict pressure wave behaviour. Therefore, we have analyzed the propagation of pressure waves in one-dimensional and two-dimensional configurations, i.e. a pipe and a slab, filled with liquid water. The pressure waves are driven by one-sided pressure boundary conditions, in the one-dimensional case of harmonic or Gaussian shape and in the two-dimensional case also of harmonic shape. The selected harmonic pressure boundary conditions lead to standing pressure waves, while using the Gaussian shape boundary conditions one-dimensional pressure pulses are injected and propagate through the pipe. The agreement of the TRACE results with the analytical solutions are, in general very good to good for the one-dimensional cases with respect to the pressure maxima and a small difference is only obtained in the wave speed. At the resonance frequencies of the one-dimensional standing waves, the code is tested to the extreme and shows that enforcing small time step sizes is crucial for the performance of the code. Non-linear effects are observed in the code results for the large amplitudes encountered at the closest neighborhood of the resonances, where the analytical linear standing wave solution diverges and the linear approximation is outside of its validity range. Also for these non-linear standing waves TRACE yields qualitatively physically correct behaviour as the pressure amplitudes are limited and a plateau is reached. For the one-dimensional pressure pulse of Gaussian shape the change of pulse amplitude and shape was analyzed in a longer system. The maximum amplitude of the pulse is slightly reduced as the pulse travels along the pipe. The effect of numerical diffusion on leading and trailing fronts is slightly asymmetric due to the donor-cell approach used in the numerical integration scheme of TRACE. The accuracy of the code is not negatively influenced by the reflections of the pulse at the boundaries of the pipe. As for the standing waves, the accuracy of the travelling pulse solution calculated by TRACE is negatively affected when the time steps are too large, while the effects of the spatial discretization are rather minor. For the case of two-dimensional standing waves in a slab, a lowest spatial harmonic generated with one wave node in the direction parallel to the pressure driving boundary is considered. TRACE results show an overall good agreement with the linear analytical solution. This good agreement includes for low to medium excitation frequencies the damping properties of the skin effect perpendicular to the pressure boundary, which does not exist in one-dimensional pressure wave propagation, the transition to a harmonic shape of the wave also perpendicular to the pressure boundary and the frequency dependence of the resonance spectrum for further increased frequencies with the rapid changes of the wavelengths encountered. Effects of the model set-up and code limitations with respect to the two-dimensional TRACE model set-up using a TRACE VESSEL component in connection with pressure boundary conditions are discussed, in particular with respect to the underestimation of the damping in the skin effect frequency range and the numerical damping for higher frequencies. All in all, the TRACE code is able to calculate one- and two-dimensional pressure wave propagation in liquid water, when an appropriate spatio-temporal numerical discretization is chosen
[en] The capabilities of the nuclear system transient codes TRACE and RELAP5 to model coupled two-phase flow and pressure wave propagations in a pipe are assessed by analyzing the UMSICHT PPP cavitation water hammer experiments 329 and 135 after valve closure. Time-dependent pressure, flow behaviour, and the generation and collapse of vapor bubbles at the valve and the first bridge are discussed. We show that both codes are able to model the flow behaviour of the water hammer for the high pressure and high temperature case 329 (initially 10-13 bar and 420 K), however condensation heat transfer for the base case needed to be increased in order to accurately model the magnitude of the first pressure excursion. The experimental broadening and damping of the subsequent pressure peaks by Fluid-Structure Interaction (FSI) phenomena arising from the interaction of the flow with the vibrations of the piping structure are not considered in the modeling results. For the lower pressure and temperature case 135 (initially 1-4 bar and 294 K), the TRACE code provides a good approximation of the propagation of the pressure wave and the void fraction behaviour, already with base case conditions, while RELAP5 overpredicts the vapor generation along the pipe and, as a result, considerably underpredicts the pressure amplitudes and overpredicts the water hammer frequency