Analysis of HDR blowdown experiment for fluid-structure interactions with multiflex

The HDR experimental facility of Kahlsruhe is comprised of a full-scale pressure vessel, core barrel, and piping systems. In the blowdown experiment, V31.1, the fluid-structure interaction of the core barrel and downcomer water is significant. This experiment is analyzed in the present paper. The HDR downcomer annulus is modeled by the one-dimensional network that is equivalent to two-dimensional fluid-structure interactions. The core barrel is modeled by the projector method for combined beam and shell models. The vessel motion is taken into account by means of the relative modal analysis proposed in this paper. Computed time histories of pressure, pressure differentials, and barrel wall displacements are compared with the experimental data. Fair agreement between experiment and post-test computation is found. Effects of the vessel motion are also discussed.     

Experimental results of coupled fluid-structure interactions during blowdown of the HDR-vessel and comparisons with pre- and post-test predictions

Experimental results of the first three blowdown tests with reactor pressure vessel internals (core barrel) at the HDR-facility in Kahl, Federal Republic of Germany, are summarized. The major goal of these and the following experiments to be performed during the first half of 1982 is the determination of the importance of multidimensional fluid-structure interaction phenomena during the initial phase of blowdown depressurization transient following a simulated pipe rupture. As such, the performed Preliminary Test Series provides the first, most realistic data in this area available thus far. The experiments have been accompanied by a series of pre- and post-test predictions with several German and American computer codes specifically developed to account for the phenomena of interest. These codes will be discussed briefly and the results presented in comparison with the data. These comparisons allow the verification of these codes on the basis of multidimensional, measured quantities for the first time. Overall conclusions on the basis of these comparisons will be presented from the point of view of Project HDR. Individual contributions by the various institutions which participated in these computational efforts supplement this information in the following articles of this special issue. The test matrix of the forthcoming Main Test Series (V31.2 through V34), defined on the basis of the foregoing experiments, concludes this overview.     

Fluid-structure analysis for the HDR blowdown and snapback experiments with FLUX

FLUX is a special purpose code to analyse three-dimensional fluid-structure interactions during blowdown of a pressurized water reactor. Such a blowdown has been simulated in the HDR experiments. For the first series of blowdown experiments and for snapback experiments in the same facility the results of precomputations are reported and compared with the experimental results. Refinements are desirable with respect to two-phase-damping of pressure waves in the blowdown pipe and vessel wall flexibility. The general quantitative agreement between measurement and computation is satisfactory.     

Analysis of the HDR blowdown experiment using a weakly coupled fluid-structural dynamic method

The methodology and results of B&W's prediction of the German HDR blowdown experiment number V31/V31.1 are presented. The test is designed as a full-scale PWR LOCA simulation. The analysis is based on a weakly coupled structural priority approach to the coupled fluid-structure problem. The hydraulic forcing function is computed first using a one-dimensional non-steady finite difference representation of the conservation equations. Structural boundaries are assumed rigid. Thus, computation of the actual pressure transient is precluded. Structural analysis is performed with a finite element lumped mass computer program. Only beam modes are considered. Fluid-structure coupling is accounted for by a separately generated hydrodynamic mass matrix. The hydraulic forcing function and the hydrodynamic mass matrix form the input basis to the structural dynamic analysis. Analytical results are described and compared favorably with measurements. The rigid boundary assumptions are shown to be conservative.     

Pre-and post-test analyses with K-FIX for an HDR blowdown test

The coupled fluid-structure dynamics of a pressurized water reactor core support barrel can be calculated with the K-FIX(3D, FLX) code for blowdown and seismic induced transients. The K-FIX solution method has been used to perform pre- and post-test analyses of a full scale blowdown test at the HDR facility in Frankfurt, Germany. The results verified the accuracy of the fully three-dimensional method, which solves the nonequilibrium, two-fluid equations for the fluid dynamics and the Timoshenko elastic shell equations for the core barrel motion.     

Homogenized model for fluid-structure interaction of the pressurized water reactor core internals during blowdown

A method for the numerical simulation of the pressurized water reactor core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. In order to investigate the global core motion during the blowdown accident, the core model describes the coupled fluid-rod motion with Homogenization methods. The heterogeneous fluid-rod mixture thus is treated as a special continuum with anisotropic material properties. Furthermore, the core model considers elastical rod forces against bending and axial straining and the direct interaction of neighbouring fuel elements, which is a highly nonlinear process due to the finite gaps. Because this effect is very important, two simulation models have been developed and are compared. All these models have been implemented into the blowdown code FLUX-4. With the new code version FLUX-5 the PWR-blowdown is parametrically investigated.     

Design of the HDR experimental program on blowdown loading and dynamic response of PWR-vessel internals

In reactor safety, a postulated breach of the primary coolant circuit, i.e. a reactor vessel blowdown, must not result in an uncontrolled failure propagation. Therefore deformations of the pressure vessel internals caused by the sudden pressure release within the first milliseconds of the blowdown must be within certain limits. To guarantee this, up to now conservative assessments are used based on theoretical models and a number of small scale experiments.In the next two years full scale blowdown experiments will be performed at the former HDR-reactor. They will be used to verify state-of-the-art and newly developed computer codes. In order to obtain reliable information on the safety margins in case of a blowdown, conservative approaches are replaced by the more detailed models of best estimate codes. Especially the coupling between fluid- and structural dynamics, i.e. the feedback of the structural deformations on the blowdown loading will be taken into account.In this paper the design of the HDR-experiments and the accompanying program of computer code development is discussed.     

Analytical method for solving fluid-structure interactions in BWR pressure suppression pool

An analytical method is developed for solving the fluid-structure interactions (FSI) associated with hydraulic transients in the BWR pressure suppression pool during a hypothetical loss-of-coolant accident. The present method assumes inviscid and infinitesimal motion of pool water for which the governing fluid-dynamic equation reduces to a linear wave equation with inhomogeneous boundary conditions of flexible pool boundaries. The present method is applied to quantitative evaluation of the FSI effect on test data obtained from a large scale pressure suppression test which was performed to investigate the pressure oscillations in the pool induced by unstationary steam condensation.     

A UK analysis of light gas distribution experiment E11.2 in the HDR facility

The HDR facility is a large-scale containment building which is highly compartmentalised except for the upper, dome region. The E11.2 experiment in HDR began with a simulated small-break LOCA, followed by a period of light gas injection. The CONTAIN 1.11 code was used in the UK to analyse this test. The predicted pressure history is shown to be in good agreement with the measurements, however, the degree of thermal and compositional stratification within the containment was underestimated. Uncertainties in instrument cooler power distribution and difficulty in precisely estimating flowpath resistance coefficients in such a complex geometry can contribute to such discrepancies. In addition, it is shown that much of the discrepancy may arise from the inability to accurately predict certain types of flows within a single physical volume with the kind of engineering flow equations used in the code.     

Vapor generation model for flashing in the initial blowdown phase

A vapor generation model for flashing in the initial blowdown phase is proposed based on a wall nucleation theory and a bubble transport model. Comparisons are made between the proposed model and the TRAC-PF1 model by using the MINCS code through analyses of three blowdown experiments with different scales. The present model well predicts the pressure undershoot in the vessel, while the TRAC model can not predict this typical thermodynamic nonequilibrium phenomenon.     

Experimental study on the blowdown load during the steam generator feedwater line break accident in the evolutionary pressurized water reactor

Integral effect tests using the ATLAS facility were performed to obtain the thermal-hydraulic parameters such as dynamic and static pressures, local temperatures, and flow rates during a feedwater line break of a steam generator. The break of a feedwater line was simulated using a double rupture disc assembly in order to satisfy the requirements for the break opening time of around a few milliseconds. In the present study, estimated break opening time was less than 1.5 ms and broken areas were 48.1% and 93.4% of the feedwater line, respectively. The maximum dynamic pressures of about 1.57 bar were obtained inside of feedwater box that was closest to the break location of the feedwater line. After the break of the feedwater line, propagation of the pressure wave along the distance from the break location inside the steam generator was clearly and pertinently observed in all the tests. From a structural integrity point of view, however, the risk induced by this maximum dynamic load could be treated to be insignificant.     

The JET ITER-like wall experiment: First results and lessons for ITER

The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.     

Integrity of the blowdown piping systems in a nuclear power plane-findings and consequences

In a pressurized-water reactor (PWR) the blowdown pipes from the four steam generators (material 1.5415) were examined. All the quality reports of the piping materials and the welding seams as strength and toughness and all non-destructive test findings of the welding seams were listed. According to pipe design drawings, a static calculation of the system under operational loads was conducted. In some cross sections adjacent to elbows an overload was estimated. Selective non-destructive testing of affected sections and inspection of the existing pipe-support conditions were recommended. Non-destructive tests during the inspection in 1994 revealed some longitudinally oriented findings in the base material in the inside walls of two elbows; the two elbows were replaced. The causes of findings were ‘in-plane bending overloading', corrosion processes in the two elbows with findings and, most probably, condensation hammers during the startup period of the plane (possibly the opening of the shutoff valves was too fast) which yielded a large displacement of the piping system. Based on findings, the following should be implemented on the 1996 inspection: (1) the two supports U 27 and U 28 are to be changed; (2) the six highly loaded elbows are to be replaced (wall-thickness is to be increased from 5.6 to 6.3 mm); (3) the whole pipework from the four steam generators to the protection barrier is to be replaced.     

Homogenisation method for the dynamic analysis of a complete nuclear steam generator with fluid-structure interaction

The present paper deals with the dynamic analysis of a steam generator tube bundle with fluid-structure interaction modelling. As the coupled fluid-structure problem involves a huge number of degrees of freedom to account for the tube displacements and the fluid pressure evolutions, classical coupled method cannot be applied for industrial studies. In the present case, the three-dimensional fluid-structure problem is solved with an homogenisation method, which has been previously exposed and successfully validated for FSI modelling in a nuclear reactor [Sigrist, J.F., Broc, D., 2007a. Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid-structure interaction coupling. Nuclear Engineering and Design 237, 431-440]. Formulation of the homogenisation method for general two- and three-dimensional cases is exposed in the paper. Application to a simplified, however representative, model of an actual industrial nuclear component (steam generator) is proposed. The problem modelling, which includes tube bundle, primary and secondary fluids and pressure vessel, is performed with an engineering finite element code in which the homogenisation technique has been implemented. From the practical point of view, the analysis highlights the major fluid-structure interaction effects on the dynamic behaviour of the steam generator; from the theoretical point of view, the study demonstrates the efficiency of the homogenisation method for periodic fluid-structure problems modelling in industrial configurations.     

Experimental and calculation results of the integral reflood test QUENCH-15 with ZIRLO™ cladding tubes in comparison with results of previous QUENCH tests

The QUENCH-15 experiment investigated the effect of ZIRLO™¹ cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (standard Zircaloy-4), QUENCH-12 (VVER, E110), and QUENCH-14 (M5^®). The QUENCH-15 bundle cross-section corresponded to a Westinghouse PWR core design and consisted of 24 heated rods (internal tungsten heaters between 0 and 1024 mm axial elevation, cladding oxidised region between −470 and 1500 mm), six corner rods made of Zircaloy-4, two corner rods made of E110, and a Zirconium 702 shroud. The test was conducted in principle with the same protocol as QUENCH-06, -12 and -14, so that the effects of the change of cladding material and bundle geometry could be more easily observed. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1473 K over a period of about 3000s. The power was then ramped at a rate of 0.25 W/s/rod to cause a temperature increase until the desired maximum bundle temperature of 2153 K was reached. The maximum oxide layer thickness observed was 380 μm. Then reflood with 1.3 g/s/rod water at room temperature was initiated. The electrical power was reduced to 175 W/rod during the reflood phase, approximating effective decay heat level. The post-test metallography of the bundle showed neither noticeable breakaway oxidation of the cladding nor melt release into space between rods. The average outer oxide layer thickness at hottest elevation of 950 mm was 620 μm (QUENCH-06: 630 μm). The molten cladding metal at hottest elevation was localised between the outer and inner oxide layers. The thickness of inner oxide layer reaches 20% of that of the outer oxide layer. The measured hydrogen release during the QUENCH-15 test was 41 g in the pre-oxidation and transient phases and 7 g in the quench phase which are comparable with those in QUENCH-06, i.e. 32 g and 4 g, respectively. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2. The calculation results support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5^® or ZIRLO™ for the beyond-design basis accident present conditions here studied.     

Experimental results on the coolability of a debris bed with multidimensional cooling effects

Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations ( [Rashid et al., 2008], [Groll et al., 2008] and [0055]). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration.A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).     

Experimental and post-test calculation results of the integral reflood test QUENCH-12 with a VVER-type bundle

The QUENCH-12 experiment was carried out to investigate the effects of VVER materials (niobium-bearing alloys) and bundle geometry on core reflood, in comparison with test QUENCH-06 using western PWR materials (Zircaloy-4) and bundle geometry. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1450 K, followed by a power ramp until a temperature of 2050 K was reached, then reflood with water at room temperature was initiated. The total hydrogen production was 58 g (QUENCH-06: 36 g), 24 g of which were released during reflood (QUENCH-06: 4 g). Reasons for the increased hydrogen production may be extensive damaging of the cladding surfaces due to the breakaway oxidation and local melt formation with subsequent melt oxidation. Post-test videoscope observations and metallographic investigations showed an influence of the breakaway oxidation with extensive spalling of oxide scales of rod claddings, shroud and auxiliary corner rods. The hydrogen content in the corner rods, withdrawn from the bundle during the test, reached more than 30 at% at the bundle elevations of 850 and 1100 mm. Post-test calculations were performed with local versions of SCDAP/RELAP5 following on from pre-test analyses with SCDAP/RELAP5 and SCDAPSIM.     

Performance of a piping system and equipment in the HDR simulated seismic experiments

This paper describes a portion of the analysis and results of the United States Nuclear Regulatory Commission/Idaho National Engineering Laboratory (USNRC/INEL) participation in the SHAG (Shakergebaude) Seismic Research Program conducted by Kernforschungszentrum Karlsruhe (KfK) at the Heissdampfreaktor (HDR), a decommissioned nuclear reactor. The program analyzed the responses of a piping system and associated line-mounted equipment when subjected to various seismic and hydraulic loadings. Of interest was to evaluate the influence that piping support system flexibility has on piping system responses. The results of the studies will contribute to the technical basis for assessing the responses of light water reactor (LWR) piping and fine-mounted equipment to earthquakes.     

Experimental and calculation results of the integral reflood test QUENCH-14 with M5® cladding tubes

The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institut (Switzerland) using the SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. Follow-on post-test analyses were performed using SCDAP/RELAP5 and MELCOR as part of an ongoing programme of model validation and code assessment. Alternative oxidation correlations were used to examine the possible influence of the M5® cladding material on hydrogen generation, in comparison with Zircaloy-4.