Post-test Calculations Research Articles

GOTHIC Verification on Behalf of the Heiss Dampf Reaktor Hydrogen-Mixing Experiments

GOTHIC is an advanced code for thermal-hydraulic flow simulation in nuclear reactor containments and other confinements. Battelle Europe has participated in the development and verification of the code over a number of years.Described are blind and open posttest calculations of the Heiss Dampf Reaktor hydrogen-mixing experiments E11.2 and E11.4 in comparison with measured data.Caused by a miscalibration of steam source flow rates and neglect of the heat sink capability of the instrumentation cooling lines, the pressure, temperature, and steam concentrations were drastically overpredicted by the blind predictions and are quantitatively not comparable to experimental data.The following conclusions can be drawn from the parametric open posttest calculations:1. Scenarios with homogeneous containment atmosphere (like E11.4) can be simulated accurately with lumped-parameter models.2. The lumped-parameter method is seemingly not fully qualified to predict the hydrogen distribution in a stratified containment atmosphere (like E11.2).3. Improved predictive quality may require the application of distributed parameter models.

Determination of Code Accuracy in Predicting Small-Break LOCA Experiment

Quantitative evaluation of thermal-hydraulic code uncertainties is a necessary step in the code assessment process, especially if best-estimate codes are utilized for licensing purposes. With the goal of quantifying code accuracy, researchers in the past developed a methodology based on the fast Fourier transform (FFT) that consisted of qualitative and quantitative code assessment. Here, the FFT-based method is applied to International Atomic Energy Agency (IAEA)-Standard Problem Exercise (SPE)-4 test results with pre- and posttest code calculations of the IAEA-SPE-4 experiment. Four system codes (ATHLET, CATHARE, MELCOR, and RELAP5) are used for calculations of the experiment, performed at the PMK-2 facility, which simulated a cold-leg break in a WER-440 plant. The results show that the posttest calculations had better accuracy than did the pretest calculations. None of the best three pre- and posttest calculations were able to predict core dryout, which was the most important phenomenon observed during the test. The results obtained can give an objective indication of the capability of the aforementioned codes in predicting relevant variables characterizing the transient (too few experimental parameters may limit full application of the FFT-based methodology).

Irregularities induced by nonstructural masonry panels in framed buildings

The effects of the nonstructural masonry infills on the global seismic behaviour of frame structures are discussed, based on the results of full-scale pseudodynamic tests and the post-test calculations. The tests were performed on a four-storey framed structure designed according to Eurocode 8, with different infill configurations. The results show that an irregular distribution of the panels yields unacceptably large damage in the frame. In addition, it is shown that even a regular distribution of infills can lead to irregular behaviour of the frame. The possibility of predicting the behaviour of such structures by means of global models is explored. Based on the comparison of the experimental and numerical results a set of guidelines for calibrating the global models to reliably predict the behaviour of this class of structures is given.

Simulation of a Small Cold-Leg-Break Experiment at the PMK-2 Test Facility Using the RELAP5 and ATHLET Codes

Results of a small-break loss-of-coolant accident experiment, conducted on the PMK-2 integral-type test facility are presented. The experiment simulated a 1% break in the cold leg of a VVER-440-type reactor. The main phenomena of the experiment are discussed, and in the case of selected events, a more detailed interpretation with the help of measured void fraction, obtained by a special measurement device, is given. Two thermohydraulic computer codes, RELAP5 and ATHLET, are used for posttest calculations. The aim of these calculations is to investigate the code capability for modeling natural circulation phenomena in VVER-440-type reactors. Therefore, the results of the experiment and both calculations are compared. Both codes predict most of the transient events well, with the exception that RELAP5 fails to predict the dryout period in the core. In the experiment, the hot- and cold-leg loop-seal clearing is accompanied by natural circulation instabilities, which can be explained by means of the ATHLET calculation.

Posttest Calculations of Bundle Quench Test CORA-13 with ATHLET-CD

Gesellschaft für Anlagen- und Reaktorsicherheit is developing, in cooperation with the Institut für Kernenergetik und Energiesysteme, Stuttgart, the system code Analysis of Thermalhydraulics of Leaks and Transients with Core Degradation (ATHLET-CD). The code consists of detailed models of the thermal hydraulics of the reactor coolant system. This thermo-fluid dynamics module is coupled with modules describing the early phase of the core degradation, like cladding deformation, oxidation and melt relocation, and the release and transport of fission products. The assessment of the code is being done by the analysis of separate effect tests, integral tests, and plant events. The code will be applied to the verification of severe accident management procedures.The out-of-pile test CORA-13 was conducted by Forschungszentrum Karlsruhe in their CORA test facility. The test consisted of two phases, a heatup phase and a quench phase. At the beginning of the quench phase, a sharp peak in the hydrogen generation rate was observed.Both phases of the test have been calculated with the system code ATHLET-CD. Special efforts have been made to simulate the heat losses and the flow distribution in the test facility and the thermal hydraulics during the quench phase. In addition to previous calculations, the material relocation and the quench phase have been modeled. The temperature increase during the heatup phase, the starting time of the temperature escalation, and the maximum temperatures have been calculated correctly. At the beginning of the quench phase, an increased hydrogen generation rate has been calculated as measured in the experiment.

10.O.04 The effect of binary condensation on post-test calculations for the integral test FAL-18

The RAFT computer code for aerosol formation and transport has been used to simulate the Falcon integral test FAL-18. We have included in RAFT the effect of condensation of a mixture of two vapours on aerosol particles and on the pipe walls. The results obtained are presented here and show that binary condensation of the CsI-CsOH system has some effect in the calculated deposition profiles of caesium. When binary condensation is taken into account the caesium deposition profile calculated by RAFT is closer to the experimental results than that calculated with the base version of the code, but still quantitatively far from the experimental results. This could be due to the complexity of the problem, which really is a case of multicomponent condensation and also to the lack of knowledge for the molal activity coefficients.

An experimental study of bottom rewetting in a rod bundle geometry

The effect of various flow parameters on the rewetting velocity in a bottom rewetting phenomena has been experimentally studied in a 3×3 rod bundle geometry. The rod bundle geometry consisted of nine test rods surrounded by a radiatively heated square shroud. The rewetting tests were conducted at atmospheric pressures for low mass and heat flow ranges. Typical rewetting velocities measured in this study were in order to 1 to 5 mm/s. For a given test condition, an increasing inlet subcooling was accompanied by an increasing rewetting velocity. The rewetting velocity was first found to be decreased with increasing vapor quality until a critical value, for a given mass flux. For the vapor qualities higher than the critical value, the rewetting velocity was found to be increased with increasing vapor quality. For a given mass flux, the critical value of the vapor quality corresponded to the flooding limit of the flow. A Transient Reactor Analysis Code (TRAC) post-test calculation was performed for a subcooled inlet reflood test. The TRAC model considered only the test bundle and consisted of four components including ten fluid cells. The calculated and measured wall and vapor temperatures were the principle parameters for comparison. The precursory cooling rate above the quench front is found to be overpredicted by TRAC, resulting in overprediction of the rewetting velocity.

The CATHARE code strategy and assessment

The French thermal-hydraulic code CATHARE is being developed in Grenoble by EDF, FRAMATOME and CEA. The best-estimate objectives of CATHARE require a very rigorous methodology for development and assessment. In particular, the validation process is performed in two steps: qualification of the models against separate effect tests and verification of the code on integral loop tests. The validation reports and the resulting evaluation document and user guidelines manual are delivered to the users. This paper illustrates the methodology for the case of CATHARE 1 version 1.3. The qualification process was conducted with reference to a matrix of 300 tests selected from experiments on critical flow, flow pattern determination, blowdown of heated or adiabatic test sections in various geometries, reflooding, boil-off, steam generator and pump behaviour, fuel thermomechanics. The results show the adequacy and the consistency of the models but indicate some items to be further studied: droplet diameter correlation, incipience of boiling model, critical flow in diaphragms, top-down rewet or reflood, interphase friction in rod bundle geometry. The verification process was performed on 21 tests coming from LOBI, LOFT, LSTF, PKL facilities. The natural circulation post-test calculations show good results for single phase or two-phase mass flowrate evaluation and a satisfactory prediction of the reflux condensor phase. The various phases of the small break or large break transients are generally well described by CATHARE. However, some quantitative discrepancies underline limited deficiencies in the domain of condensation at safety injection, CCFL at steam generator inlet or core upper tie plate, multidimensional effects in annular downcomer and in core region, modelling of steam generator secondary side. The consideration of these items led to the introduction of significant improvements and new capabilities in the CATHARE 2 version which consequently has to be validated following the same two-step process and with an extended test matrix. The efficiency of such a rigorous methodology has induced the French nuclear industry (EDF, FRAMATOME) to use CATHARE for safety research and development studies.

Piping analysis of large scale experiments with adina and dapsy

Large scale piping experiments have been carried out recently in Germany at the HDR test facility. The aim of these tests was to investigate the fluid and structural dynamic behaviour of a feed-water line and its check valve. GRS has carried out pre- and post-test calculations of these experiments using the fluid dynamic code DAPSY and the structural analysis code ADINA. Staue and modal analysis results with ADINA and both coupled and uncoupled analysis results of pressure transients with ADINA and DAPSY are presented and compared with experimental data. The main finding of the work presented in the paper is that the highly sophisticated pipe element in ADINA gives adequate results at reasonable costs and is particularly suitable to investigate special features such as the ovalization of the pipe or local stresses.

Stresses and strains in the steel containment resulting from transient pressure and temperature loading during loss-of-coolant accident

Abstract Posttest calculations of stresses and strains in the steel containment of the German research reactor HDR were performed for a simulated LOCA. The results of the theoretical investigations will be presented and compared to experimental findings. The pressure and temperature loading of the shell was determined with the thermodynamic code COFLOW on the basis of a multi-compartment model. Using a three-dimensional finite element model the temporal behaviour of the containment was calculated employing the structural mechanics code ASKA. Global bending deformations and local negative straining of the steel shell is discussed. Theoretical and experimental results agree in most cases rather well. Reasons for deviations will be discussed. The specific behaviour of strains found in the vicinity of locally heated areas will be explained by means of analytical considerations.

Hydrodynamic loading and response of reactor internals using 3D STEALTH/WHAMSE

A three-dimensional methodology was developed for realistically predicting fluid-structure interaction transient loads and the structural response of the rector vessel, core support barrel, and core during the subcooled portion of a hypothetical loss-of-coolant accident. As described in this paper, the methodology uses a hydrodynamics computer program, STEALTH 3D, coupled with a structural response program WHAMSE 3D, to calculate the hydrodynamic and structural behavior of two tests in the HDR blowdown program in Karlstein, Germany. The posttest calculation of one test and the pretest calculation of the other are compared with the experimental data from the two tests, and reasonably good agreement is observed. The methodology was then applied to the calculation of the structural response of simulated pressure vessel internals during a postulated LOCA of a PWR. The model was a modification of the model in one of the HDR calculations. The PWR/HDR model included a simplified core represented by an axially located center beam so that the configuration was more typical of a large PWR than was the HDR model. Results of the PWR/HDR calculation are compared with the HDR results. Transient response of the axially distributed beam core is also discussed.

The Application of Listen-Read-Discuss (L-R-D) Strategy to Improve Reading Skill at the First Grade of SMPN 10 Parepare

This study aim to find out the difference in reading skill between students who are taught of using L-R-D strategy and without using L-R-D strategy at the first grade of SMPN 10 Parepare. Furthermore, its aim is to determine the effectiveness of the Listen-Read-discuss strategy in reading skill learning at the first grade of Junior High School 10 Parepare. This research is an experiment design using pretest posttest design. The population of this research was all classes VII of SMP Negeri 10 Parepare, with samples of class VII 4 was experiment class and class VII 3 was control class. Data was obtained by Barret’s taxonomy reading skill test. The validity used is the validity of items that are assisted by computer item program. Reliability using the alpha Cronbach formula obtained result of 0.93. the data analysis technique used T-test with a significance of 5%. Before the data analysis was carried out using t-test technique, the prerequisite analysis test was carried out in the form of data distribution normality test and variant homogeneity test. The results of the pretest and posttest calculation of both groups are normal and homogeneous. T-test analysis of pretest data in reading skill of the experimental group and the control group, obtained -0.930 and (sig 0.357>0.05). This result indicates the initial skill of the experimental group and the control group were the same or there were no significant differences between the two group. T-test result of posttest in reading skill of the experimental group and the control group obtained -3.505 and (sig 0.000< 0.05) based on these data the results of this research showed significant differences between groups taught using of Listen-read-discuss strategy and without Listen-read-discuss strategy. The result of analysis T-test of the pretest and posttest showed that Listen-read-discuss strategy was effective to improve reading skill at the first grade of SMPN 10 Parepare.