Thermal-hydraulic Code ATHLET Research Articles

Experimental and theoretical investigation of the boiling heat transfer in a low-pressure natural circulation system

The implementation of passive safety systems in nuclear reactors provides the opportunity to enhance the nuclear safety. On the other hand, an accurate and reliable prediction of the heat removal behavior is not ensured because the operating conditions of certain types of passive systems like containment cooling systems differ from the validity ranges of the established heat transfer correlations. Therefore, a generic and detailed investigation is still necessary for passive systems. Against this background, the test facility GENEVA was erected at Technische Universitat Dresden in 2012. Since the commissioning, generic experiments concerning the system and stability behavior of this facility, which emulate a low-pressure and low-flow (LPLF) natural circulation system, were provided. Nevertheless, the investigation of the heat transfer behavior remained an open issue. On this account, the instrumentation in the heat transfer region inside GENEVA was improved to gather the necessary temperature and void fraction profiles. The performed experiments provide a generic and wide database concerning boiling in a LPLF natural circulation systems. Within this paper, the development of the wall and bulk fluid temperature as well as the axial and center line void fraction profile in a slightly inclined tube for different heat flow rates are discussed. Furthermore, flow patterns could be identified on behalf of the void fraction measurements. To conclude the experimental analysis, the development of the heat transfer coefficient was estimated. These experimental data provide the basis for a simulation with the lumped-parameter thermal-hydraulic code ATHLET and serve as validation reference. However, the comparisons between the experimental and computational results show insufficient agreements. Mainly, the simulation misses the saturation point of the experiments, which leads to great differences of the void fraction values. Moreover, inaccuracies appear as well with the heat transfer coefficient. The experimental and computational results that are discussed in this paper provide the basis for the advancement not only of heat transfer correlations but also of flow pattern maps within the range of low pressure natural circulation system. In summary, this investigation contributes to the general purpose to enhance nuclear safety by providing an accurate and reliable prediction o f the heat removal capacity of passive systems.

ATHLET-CD/COCOSYS Analyses of Severe Accidents in Fukushima Daiichi Units 2 and 3: German Contribution to the OECD/NEA BSAF Project, Phase 1

On behalf of the German Federal Ministry of Economics and Technology, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) participated in the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) project titled Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF). Analysis of the severe accidents (SAs) that happened in the Fukushima Daiichi nuclear power plant (NPP) requires well-qualified methods and codes, e.g., ATHLET-CD and COCOSYS developed and applied at GRS. Coupled ATHLET-CD/COCOSYS analyses for the SA progression during the first days for the similar Units 2 and 3 of Fukushima Daiichi have been provided as the German contribution to the OECD/NEA BSAF project, phase 1. ATHLET-CD is a detailed SA code based on the thermal-hydraulic code ATHLET of GRS to simulate the processes in the reactor circuit before and during core degradation. COCOSYS is focused on the simulation of design basis and SA progression in the containment and the surrounding buildings of the NPP.The focus is on selected results of the SA analyses in the boiling water reactors at the Fukushima Daiichi site especially with regard to the conditions in the torus-shaped wetwell (WW) of the primary containment and specific modeling needs. The GRS results obtained in this OECD/NEA BSAF project, phase 1, are encouraging in terms of capturing essential SA signatures like reactor and containment pressure, reactor water level, and WW temperature history for the first days of the accident in the analyzed Units 2 and 3. A detailed plant model was built up especially with a detailed torus nodalization allowing modeling of relevant phenomena like thermal stratification in the torus water pool and consideration of plant-specific details with regard to local water/steam injections into the torus water pool through safety systems and valves. As a result, the calculated accident progression of the best-estimate analyses for both units follows the accident time line quite closely. This is a prerequisite for reasonable core degradation calculations, as the time window available for the onset of core degradation between known points in time when safety injection stops and mobile pump injection into the reactor starts is small. The analyses are useful to identify areas that require further attention, to define information needs to be gained from the decommissioning, and to define further research needs with regard to experiments and code improvement.

Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena.Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.

Validation and verification of the coupled neutron kinetic/thermal hydraulic system code DYN3D/ATHLET

One of the most intensively developing areas in the LWR multi-physics is a coupling of different best estimate 3-D neutron kinetic (BIPR, DYN3D, KIKO3D, NEM, PARCS, etc.) and thermal hydraulic (ATHLET, CATHARE, RELAP5, etc.) codes. Resulting coupled code systems have advanced capabilities of modeling both steady-state spatial distributions of the core power and their evolutions during different kinds of reactor transients. They are also highly useful in the analyses of possible reactor instabilities. Initial steady-state core power distributions can be disturbed by changes in the reactor loop mass flow rates and/or temperatures, by relocations of the low-temperature/diluted-boron water slugs within the primary system or by movements of control rods.The coupled code used for LWR simulations in HZDR is DYN3D/ATHLET, which includes the 3-D core neutron kinetic and thermal hydraulic model of own development – DYN3D. The paper reports major capabilities of DYN3D as well as different ways of its coupling with the thermal hydraulic code ATHLET (external, internal and parallel), but mainly focuses on the validation and verification of the coupled code DYN3D/ATHLET. In the course of DYN3D/ATHLET validation/verification nearly 20 real plant transients and dynamic benchmarks have been simulated and analysed. The LWR types covered by these tasks are: VVER-440 (Bohunice-3, Greifswald-5 and Loviisa-1 units), VVER-1000 (Balakovo-1, Balakovo-4, Kalinin-3, Kozloduy-6, Saporozhye-6 and Temelin-2 units), B&W PWR (TMI-1) and BWR/4 MK-1 (Peach Bottom-2).For each reactor unit a computational model was developed according with the benchmark specifications. The simulated tasks describe different scenarios of increasing complexity, including the transients initiated by the main steam header or main steam line breaks, switching off/on of main circulation pumps, turbine trip and generator load drop. Some of the transients are characterized by a strongly asymmetric behavior of the primary system (e.g. caused by a steam line break), and the processes of coolant mixing in the lower and upper reactor plenums as well as in the downcomer are important for these cases. The coupled code DYN3D/ATHLET models the primary coolant mixing in two ways – whether by an appropriate nodalization of the mixing area or by using two specific models for mixing in the lower plenum. The first of the lower plenum mixing models is a part of the coupling interface at the core inlet plane, while the second one is the analytical coolant mixing model developed for the downcomer and the lower plenum regions of VVER-440.The results of DYN3D/ATHLET simulations were assessed both against available measured data and calculations performed with similar multi-physics codes. The paper includes an overview of the simulated problems and the most representative results of DYN3D/ATHLET validation/verification for all coupling modes. The obtained experience of code validation provides a better understanding of reactor transients with a strong interaction between neutron kinetics and thermal hydraulics, helping to improve computation models. This experience is also useful for the present and further activities in coupling of DYN3D with other best estimate codes, like CFX, TRANSURANUS, etc.

Comparison of different coupling CFD–STH approaches for pre-test analysis of a TALL-3D experiment

The system thermal-hydraulic (STH) code ATHLET was coupled with the commercial 3D computational fluid dynamics (CFD) software package ANSYS CFX to improve ATHLET simulation capabilities for flows with pronounced 3D phenomena such as flow mixing and thermal stratification. Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), validation activities for coupled thermal-hydraulic codes are being carried out. The TALL-3D experimental facility, operated by KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant at natural and forced circulation conditions. GRS carried out pre-test simulations with ATHLET–ANSYS CFX for the TALL-3D experiment T01, while KTH scientists perform these analyses with the coupled code RELAP5/STAR CCM+. In the experiment T01 the main circulation pump is stopped, which leads to interesting thermal-hydraulic transient with local 3D phenomena. In this paper, the TALL-3D behavior during T01 is analyzed and the results of the coupled pre-test calculations, performed by GRS (ATHLET–ANSYS CFX) and KTH (RELAP5/STAR CCM+) are directly compared.

Improvement of ATHLET modelling capability for asymmetric natural circulation phenomenon using uncertainty and sensitivity measures

Natural circulation is one of the most important phenomena relating to nuclear reactor safety. In the last years of 20th century, lots of test facilities were set up to conduct natural circulation experiments. At the same time, some system codes, such as RELAP5 and CATHARE, have been validated by the experimental data carried out by these facilities. The calculation results show that these codes can capture most of the characters during the natural circulation transient, except some special processes. One of the examples is PKL IIIB3.1 test, an asymmetric natural circulation transient in the primary loop with an isolated steam generator. The cooled down process using the intact steam generators by reducing the secondary side pressure is the driving force of the natural circulation in primary loop.In this paper, a post-test calculation is performed by best-estimate thermal-hydraulic code ATHLET to simulate PKL IIIB3.1 test. Due to the effect of isolated steam generator on the whole primary cooling system, the results show that the mass flow in the isolated loop is overpredicted significantly comparing to the experimental data. Many measures have been tried to improve the simulation results, but an effective method is still missing. Based on the post-test calculation results, an uncertainty and sensitivity analysis is carried out to find out the main reason causing this higher circulation mass flow. Lots of factors are considered in this study, e.g. the boundary conditions, thermal-hydraulic correlations, and model parameters. The most sensitive factors are selected and adopted to improve the simulation modelling. The results achieved so far indicate that with these measures, the ATHLET capability for the asymmetric natural circulation phenomenon is effectively improved.

Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations

Natural circulation characteristics at low pressure/low power have been studied by performing experimental investigations and numerical simulations. The PANDA large-scale facility was used to provide valuable, high quality data on natural circulation characteristics as a function of several parameters and for a wide range of operating conditions. The new experimental data allow for testing and improving the capabilities of the thermal-hydraulic computer codes to be used for treating natural circulation loops in a range with increased attention. This paper presents a synthesis of a part of the results obtained within the EU-Project NACUSP “natural circulation and stability performance of boiling water reactors.” It does so by using the experimental results produced in PANDA and by showing some examples of numerical simulations performed with the thermal-hydraulic code ATHLET.

Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam–argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B 4C oxidation do not affect significantly the total calculated hydrogen release rates.

05/01819 Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET: Hainoun, A. and Alissa, S. Nuclear Engineering and Design, 2004, 235,(1), 33–52

05/01819 Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET: Hainoun, A. and Alissa, S. Nuclear Engineering and Design, 2004, 235,(1), 33–52

Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.

Natural circulation and stability performance of BWRs (NACUSP)

From the beginning of BWR technology it was realized that a BWR can become unstable under particular circumstances caused by a feedback between the thermal-hydraulics and the neutronics. This instability can result in oscillations of the power and the flow rate, which is an unwanted phenomenon. The NACUSP project addresses the stability issues in current and future BWRs by expanding the basic understanding through well structured testing and analyses of experimental data, by analyses of existing operational stability data from three different European reactors (Forsmark, Leibstadt, Cofrentes), by applying this knowledge via efficient models and validated computer codes to operating reactors and reactor designs, and by developing general guidelines for reactor operation and design on how to avoid BWR instabilities. In order to cover the parameter range as efficiently as possible, four existing, sophisticated thermohydraulic test facilities (CLOTAIRE [Gouirand, J.M., 1988. CLOTAIRE Program, description and manufacturing of the mock-up, CEA Cadarache, DRE/STRE/LGV 88–876.] DESIRE [van de Graaf, R., van der Hagen, T.H.J.J., Mudde, R.F., 1994. Two-phase flow scaling laws for a simulated BWR assembly. Nucl. Eng. Des. 148, 455–462.] CIRCUS [de Kruijf, W.J.M., van der Hagen, T.H.J.J., Mudde, R.F., 2000. CIRCUS; a natural circulation two-phase flow facility, Eurotherm Seminar No. 63, 6–8 September 1999 Genoa, Italy, 391–395] and PANDA [Dreier, J., Huggenberger, M., Aubert, C., Bandurski, T., Fischer, O., Healzer, J., Lomperski, S., Strassberger, H.-J., Varadi, G., Yadigaroglu, G., 1996. The PANDA facility and first test results, Kerntechnik 61, 214–222]) have been selected. To extrapolate from small-scale separate-effect testing conditions to full-scale integral reactor conditions one needs to rely on the performance of computer codes (MONA [Hoyer, N., 1994. MONA, a 7-Equation Transient two-phase flow model for LWR dynamics, Proceedings of the International Conference on New Trends in Nuclear System Thermohydraulics, pp. 271–280], ATHLET [Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nucl. Technol. 128, 75–86], RAMONA-3(-5) [Grandi, G., Hoyer, N., Belblidia, L., 1998. Two-fluid thermal hydraulics capabilities of RAMONA-5, Proceedings of the International Conference on the Physics of Nuclear Science and Technology, October 5–8, 1998, Long Island, NY, USA, 1621–1625], LAPUR-V [Otaduy, P.J., March-Leuba, J., 1990. LAPUR User's Guide, NUREG/CR-542, ORNL/TM-11285], DWOS, FLICA [Toumi, I., Bergeron, A., Gallo, D., Royer, E., Caruge, D., 2000. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications, Nucl. Eng. Des. 200, 139–155], SAPHYR, RELAP5/MOD3 [Idaho National Engineering Laboratory, 1995. RELAP5/MOD3 Code Manual, Vols. 1–6, INEL-95/0174, NUREG/CR-5535], TRAC-BF1 [Idaho National Engineering Laboratory, 1992. TRAC-BF1/MOD1; an advanced best-estimate computer program for BWR accident analysis, Vols. 1–3, EGG-2626, NUREG/CR-4356]. For specific items CFD codes are applied as well. Within NACUSP a linear stability analysis tool is developed. Four of the three experimental facilities within the project have yielded a large, unique database. Natural-circulation and stability characteristics at nominal pressure were collected from extensive experiments at the DESIRE facility. Low-pressure characteristics were measured at the PANDA (large scale) and the CIRCUS (small scale) facility. The phenomena encountered (flow rate and stability trends, the occurrence of flashing induced oscillations) can be explained from basic physical models and are now well understood. Several thermal-hydraulic codes have been benchmarked—some successfully, others with limited success—against these data. The CLOTAIRE (nominal pressure, large scale) facility has been modified and is now ready for experiments. Nuclear power plant data from three BWRs (Cofrentes, Forsmark and Leibstadt) have been collected. A very complete set of reactor data was obtained from measurements at the Leibstadt BWR during cycle-19. During this test, which was performed within the NACUSP project, safe reactor operation was demonstrated at extreme conditions, covering the entire power-flow map. State-of-the-art, coupled thermal-hydraulic—neutronics codes are being benchmarked on these data. The development of an easy to use, rapid and efficient analytical tool for parametric studies on BWR stability is in progress. The last year of NACUSP will be devoted to finalising the before mentioned activities and to use the tools developed and the experience obtained for developing general guidelines for designing and operating BWRs.

Analyses of the OECD Main Steam Line Break Benchmark with the DYN3D and ATHLET Codes

The Organization for Economic Cooperation and Development (OECD) Main Steam Line Break (MSLB) Benchmark was defined to validate the thermal-hydraulic system codes coupled with three-dimensional (3-D) neutron kinetic codes. The reference problem is an MSLB in a pressurized water reactor at end of cycle. The analyses were performed with the 3-D core model DYN3D, the thermal-hydraulic system code ATHLET, and the coupled code DYN3D/ATHLET. The results of the DYN3D and ATHLET simulations based on the specification are compared with the results of other participants in the final OECD reports. The effect of the thermal-hydraulic nodalization of the core, i.e., the number of coolant channels, and the influence of the coolant mixing inside the pressure vessel are studied in the paper. Calculations with a reduced number of coolant channels are performed often in coupled calculations for saving computational time. Results of a 25-channel model were compared with the 177-channel calculation (1 channel per assembly). The results for global parameters like nuclear power show only small differences for the two models; however, the prediction of local parameters such as maximum fuel temperatures requires a detailed thermal-hydraulic modeling. The effect of different coolant mixing within the reactor pressure vessel is investigated. It is shown that the influence of coolant mixing mitigates the accident consequences when 3-D neutron kinetics is applied. In case of point kinetics, coolant mixing leads to an opposite effect. To profit from the 3-D core model, a realistic description of the coolant mixing in the coupled codes is a topic of further investigations.

Heat transfer in horizontal tubes during two phase natural circulation with presence of noncondensing gas

The condensation process of steam inside horizontal tubes during natural circulation gains in importance regarding the reactor safety research for existing and future nuclear power plants. Experimental investigations due to the condensation process were realized with the rig HORUS to study the behaviour of water-steam-gas mixtures in horizontal tubes. The paper includes statements regarding the flow and heat transfer conditions inside the tube and the temperature distribution inside the small tube wall. The experiments showed a blockade of the heat transfer area with Nitrogen which is connected with an increasing primary pressure followed by a compression of the Nitrogen and a reentry of steam into the tube. The experiments serve for the creation of an experimental data base. A model development for calculation of the heat transfer is described. The model was implemented in the German thermal-hydraulic code ATHLET. The comparison of calculated data and the measured parameters of HORUS rig show the code improvement for the simulation of noncondensing gases.

Numerical Simulation of the Emergency Condenser of the SWR-1000

The SWR-1000 is a new innovative boiling water reactor (BWR) concept, which was developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g., four emergency condensers, four building condensers, eight passive pressure pulse transmitters, and six gravity-driven core-flooding lines). In the framework of the BWR Physics and Thermohydraulic Complementary Action to the European Union BWR Research and Development Cluster, emergency condenser tests were performed by Forschungszentrum Jülich at the NOKO test facility. Posttest calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode. The one-dimensional thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of conventional finite difference calculations using the code CFX-4 are presented, which investigate the natural convection during the heatup process at the secondary side of the NOKO test facility.

Pre- and post-test calculations to natural circulation experiments at the integral test facility ISB-VVER using the thermalhydraulic code ATHLET

In 1995 at the integral test facility ISB-VVER in Elektrogorsk near Moscow natural circulation experiments were performed, which were scientifically accompanied by the Forschungszentrum Rossendorf. These experiments were the first of this kind at a test facility, which models VVER-1000 thermalhydraulics. Using the code ATHLET which is being developed by ‘Gesellschaft für Anlagen und Reaktorsicherheit’, pre- and post-test calculations were done to determine the thermalhydraulic events to be expected and to define and tune the boundary conditions of the test. The conditions found for natural circulation instabilities and cold leg loop seal clearing could be confirmed by the tests. Besides the thermalhydraulic standard measuring system, the facility was equipped with needle shaped conductivity probes for measuring the local void fractions.

Numerical simulation of the emergency condenser of the SWR1000 / Numerische Sımulation des Notkondensators des SWR1000

The SWR1000 is a new innovative boiling water reactor concept, which is developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g. four emergency condensers, four building condensers, eight passive pressure pulse transmitters, six gravity-driven core flooding lines). In the framework of BWR Physics and Thermohydraulic Complementary Action (BWR-CA) to the EU BWR R&D Cluster emergency condenser tests were performed by Forschungszentrum Julich at the NOKO test facility. In this paper post test calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode. The 1D thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of CFD-calculations using the code CFX-4 are presented, which investigate the natural convection during the heat up process at the secondary side of the NOKO test facility.

Post test calculations to 11% break loca experiments on the integral test facility ISB-VVER using the thermalhydraulic code ATHLET

The considered test was a break on the upper plenum with different modes of emergency core cooling. The reference case was the non-availability of emergency cooling. Injecting the emergency coolant into the cold leg, no increasing of rod cladding temperatures was observed, but natural circulation instabilities occurred. Injecting the cooling into the hot leg, the cooling situation was getting worse. Due to the injected cold emergency coolant, the fluid density in the discharge volume was enhanced and the break mass flow increased. The observed events in the test were reproduced by the code with good agreement.

Fast and accurate water and steam properties programs for two-phase flow calculations

The athlet code (M.J. Burwell et al., The thermalhydraulic code athlet for analysis of PWR and BWR systems, NURETH-4, Karlsruhe, 1989) is being developed at GRS for the thermohydraulic analysis of the behavior of PWRs and BWRs under postulated transient and accidental conditions. It is used by a number of institutions to perform thermohydraulic analysis in the field of nuclear reactor safety. These analyses are carried out both on mainframes and workstations. In general, these analyses consume hours of computing time on high speed computers. Much of this computing time is used to calculate water and steam properties. It has been found that, in some applications, up to 50% of the computing time is used to evaluate water and steam property routines, depending on the simulation model. In order to cut down computing costs most thermohydraulic codes use specially designed water and steam property packages which run much faster than the very accurate reference package IFC85 by Haar, Gallagher and Kell (NBS/NRC Wasserdampftafeln, London, 1988). At GRS work has been carried out to improve the packages used in the athlet code, which are fast running specially designed subroutines based on cubic and bicubic interpolation. In the reanalysis of the problem of constructing fast and accurate property packages it was found that, by using a sophisticated design of property package, substantial improvement can be achieved. The main points of the new package are as follows. 1. (i) For saturation values depending on pressure: extended range for pressure and temperature (0.01 bar, 6.9°C to 215 bar, 371.5°C); improved accuracy at the same computing effort by optimum selection of nodes; spline-based cubic interpolation leading to continuous second derivatives; common grid for all properties; consistent derivatives calculated by analytical differentiation of the interpolation functions 2. (i) For property values depending on pressure and temperature (subcooled, superheated, coexisting phases): extended range for pressure and temperature (liquid 0.01 bar, 6.9°C to 215 bar, 371.5°C; vapor 0.01 bar, 6.9°C to 215 bar, 2000°C); fast computation of subcooled and superheated values by combining analytical functions and bicubic interpolation; bicubic interpolation-based and 2D spline interpolation on a transformed curvilinear grid; linear continuation into the domain of coexisting phases (thermodynamic nonequilibrium); 2D functions to match the saturation values at the saturation line exactly; common 2D grid for all property functions; consistent derivatives calculated by analytical differentiation of the interpolation functions. The new package calculates only the material properties and leaves aside viscosity, heat capacity, thermal conductivity and surface tension, for which a fast solution is already given by Haar, Gallagher and Kell. The new package has been implemented into the athlet code. Its performance is demonstrated with the help of a typical thermohydraulic problem of a transient fill-up of a vessel. In addition, many numerical recipes and useful insights gained during the development of the new package will be presented, providing useful information for anyone who is planning to replace property packages, complicated correlations, etc. by fast and accurate interpolations/approximations.

Analysis of a selected two-phase flow phenomenon in VVER reactors with horizontal steam generators

Since 1984 the thermal-hydraulic code ATHLET has been also applied for the analyses of LOCA and transients in VVER plants. The specific design of these plants especially of the steam generator design requires a specific modeling of the phenomena which may occur under LOCA and transient conditions in these plants. Differences in design compared to the design of western reactors have been briefly listed. Specific phenomena occuring under small leak accidents are shortly described. The consideration of the simulation of the boiler-condensor mode illustrates the modelling requirements for a code which may be applied to the prediction of such a thermalhydraulic behaviour. Facing the lack of experimental data, the reliability of the simulation has been discussed by means of plausibility studies based on the momentum balance for steam and water.

International activities for the analysis of the TMI-2 accident with special consideration of ATHLET calculations

Several OECD countries still have great interest to analyze the TMI-2 accident. Thermal hydraulic best estimate codes and severe accident codes are used to calculate the TMI-2 analysis exercise defined by a CSNI task group. Fourteen organizations in nine OECD countries are participating in the exercise. Four thermal hydraulic best estimate codes and six severe accident codes are used. The Federal Republic of Germany (FRG) is using the thermal hydraulic code ATHLET developed in the GRS to calculate the TMI-2 analysis exercise. Lessons learned are concentrated on the assessment of ATHLET, show advantages of the two phase thermal hydraulic model used, and identify areas for further development. Results from ATHLET calculations are compared with results from other OECD-codes.