ASTEC: An Integral Code for Simulation of Severe Light Water Reactor Accidents

The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. Thus, the main fields of ASTEC application are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 studies as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses, as well as a combination of multiple modules for coupled effects testing and integral analyses. Among activities concentrating on the validation of individual ASTEC modules describing specific phenomena, the applicability to reactor cases marks an important step in the development of the code. Feasibility studies on plant applications have been performed for several reactor types such as the German Konvoi PWR 1300, the French PWR 900, and the Russian VVER-1000 and −440 with sequences like station blackout, small- or medium-break loss-of-coolant accident, and loss-of-feedwater transients. Subject of this paper is a short overview on the ASTEC code system and its current status with view to the application to severe accidents sequences at several PWRs, exemplified by selected calculations.

Experiment and Numerical Simulation of Bubble Behaviors in Argon Gas Injection Into Lead-Bismuth Pool

In a lead-bismuth alloy (45%Pb-55%Bi) cooled direct contact boiling water fast reactor (PBWFR), steam can be produced by direct contact of feed water with primary Pb-Bi coolant in the upper core plenum, and Pb-Bi coolant can be circulated by buoyancy forces of steam bubbles. As a basic study to investigate the two-phase flow characteristics in the chimneys of PBWFR, a two-dimensional two-phase flow was simulated by injecting argon gas into Pb-Bi pool in a rectangular vessel (400mm in length, 1500mm in height, 50mm in width), and bubble behaviors were investigated experimentally. Bubble sizes, bubble rising velocities and void fractions were measured using void probes. Argon gas was injected through five nozzles of 4mm in diameter into Pb-Bi at two locations. The experimental conditions are the pressure of atmospheric pressure, Pb-Bi temperatures of 443K, and the flow rate of injection Ar gas is 10, 20, and 30 NL/min. The measured bubble rising velocities were distributed in the range from 1 to 3 m/s. The average velocity was about 0.6 m/s. The measured bubble chord lengths were distributed from 1mm up to 30mm. The average chord length was about 7mm. An analysis was performed by two-dimensional and two-fluid model. The experimental results were compared with the analytical results to evaluate the validity of the analytical model. Although large diameter bubbles were observed in the experiment, the drag force model for spherical bubbles performed better for simulation of the experimental result because of high surface tension force of Pb-Bi.

Experiments on Gas Mixing and Stratification Driven by Jets and Plumes in Large-Scale, Multi-Compartment Geometries

The present paper discusses experiments carried out to examine mixing of different gases (steam, air) and the evolution their distributions in large-scale, multi compartment geometry imitating nuclear reactor containment compartments. The flow and the mixing process in the experiments are driven by plumes and jets representing source structures with different momentum-to-buoyancy strength. The time evolution of the relevant parameters like gas concentrations, velocities and temperatures are followed using dedicated instrumentation. The data obtained is meant to be used for the validation and development of high-resolution, mainly CFD based, 3D computational tools for nuclear reactor containment safety analysis.

Natural Convection Heat Transfer in a Rectangular Liquid Metal Pool With Bottom Heating and Top Cooling

An experimental study is performed to investigate the natural convection heat transfer characteristics with subcooled coolant to create engineering database for basic applications in a lead alloy cooled reactor. Tests are performed in the ALTOS (Applied Liquid-metal Thermal Operation Study) apparatus as part of MITHOS (Metal Integrated Thermo Hydrodynamic Operation System). A relationship is determined between the Nusselt number Nu and the Rayleigh number Ra in the liquid metal rectangular pool. Results are compared with correlations and experimental data in the literature. Given the similar Ra condition, the present test results for Nu of the liquid metal pool with top subcooling are found to be similar to those predicted by the existing correlations or experiments. The current test results are utilized to develop natural convection heat transfer correlations applicable to low Prandtl number Pr fluids that are heated from below and cooled by the external coolant above. Results from this study are slated to be used in designing BORIS (Battery Optimized Reactor Integral System), a small lead cooled modular fast reactor for deployment at remote sites cycled with MOBIS (Modular Optimized Brayton Integral System) for electricity generation, tied with NAVIS (Naval Application Vessel Integral System) for ship propulsion, joined with THAIS (Thermochemical Hydrogen Acquisition Integral System) for hydrogen production, and coupled with DORIS (Desalination Optimized Reactor Integral System) for seawater desalination. Tests are performed with Wood’s metal (Pb-Bi-Sn-Cd) filling a rectangular pool whose lower surface is heated and upper surface cooled by forced convection of water. The test section is 20 cm long, 11.3 cm high and 15 cm wide. The simulant has a melting temperature of 78°C. The constant temperature and heat flux condition was realized for the bottom heating once the steady state had been met. The test parameters include the heated bottom surface temperature of the liquid metal pool, the input power to the bottom surface of the section, and the coolant temperature.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

Measurements of Sheath Temperature Profiles in Bruce LVRF Bundles Under Post-Dryout Heat Transfer Conditions in Freon

An experimental program has been completed to study the behaviour of sheath wall temperatures in the Bruce Power Station Low Void Reactivity Fuel (shortened hereafter to Bruce LVRF) bundles under post-dryout (PDO) heat-transfer conditions. The experiment was conducted with an electrically heated simulator of a string of nine Bruce LVRF bundles, installed in the MR-3 Freon heat transfer loop at the Chalk River Laboratories (CRL), Atomic Energy of Canada Limited (AECL). The loop used Freon R-134a as a coolant to simulate typical flow conditions in CANDU® nuclear power stations. The simulator had an axially uniform heat flux profile. Two radial heat flux profiles were tested: a fresh Bruce LVRF profile and a fresh natural uranium (NU) profile. For a given set of flow conditions, the channel power was set above the critical power to achieve dryout, while heater-element wall temperatures were recorded at various overpower levels using sliding thermocouples. The maximum experimental overpower achieved was 64%. For the conditions tested, the results showed that initial dryout occurred at an inner-ring element at low flows and an outer-ring element facing internal subchannels at high flows. Dry-patches (regions of dryout) spread with increasing channel power; maximum wall temperatures were observed at the downstream end of the simulator, and immediately upstream of the mid-bundle spacer plane. In general, maximum wall temperatures were observed at the outer-ring elements facing the internal subchannels. The maximum water-equivalent temperature obtained in the test, at an overpower level of 64%, was significantly below the acceptable maximum temperature, indicating that the integrity of the Bruce LVRF will be maintained at PDO conditions. Therefore, the Bruce LVRF exhibits good PDO heat transfer performance.

Analytical Evaluation of Unsteady Thermal Hydraulic Characteristics in Shell and Tube Heat Exchangers

Many shell & tube heat exchangers are used in nuclear power plants. Unsteady thermal hydraulic phenomena have been studied in shell & tube heat exchangers to improve their safety and reliability and to extend their lifetime based on experience obtained from long periods of plant operation. We investigated unsteady flow in shell & tube heat exchangers by using computational fluid dynamics (CFD) analyses. The inlet flow on the shell side was separated and flow in several directions. A large part of the flow crossed over the tube bundle, and some parts of the flow took two circuitous roots (up and down) along the inner surface of the shell. Separated circuitous flows collided again where a baffle plate had been cut off. A pair of symmetric vortexes could be seen in that location. Some parts of the circuitous flow moved backwards into the tube bundle due to vortexes. These vortexes were unstable and changed their size and location. A pair of vortexes changed from symmetric to asymmetric. As a result, the direction of flow in the tube bundle near the vortexes changed continuously. Variations in vortexes simulated through CFD analyses could also be seen in tests on the actual size. Fluid temperature fluctuations around tubes were also evaluated through CFD analyses. Unsteady phenomena with changes from symmetric to asymmetric vortexes could be observed in the shell & tube heat exchanger and were simulated through CFD analyses with a detached eddy simulation (DES) turbulence model.

An Investigation of Flow, Heat Transfer Characteristic of Annular Flow and Critical Heat Flux in Vertical Upward Round Tube

The term annular flow is used to describe the configuration of vapor-liquid flow in which part of the liquid travels as a film on the wall and the rest is entrained as drops by the vapor core in the center of the channel. The objective of this paper is to develop a hydrodynamic model for vertical upward annular flow. A separated flow model is developed and the conservations of Mass, Momentum, Energy, entrainment rate correlation in wide range of conditions and interfacial frictional correlation are used to research the flow and heat transfer characteristic of annular flow. The liquid film thickness, liquid film mass flow rate, two-phase heat transfer coefficient pressure along axial position, local velocity profiles along radial position are predicted theoretically. The influence of the mass flux, heat flux on liquid film thickness, heat transfer coefficient etc. are investigated in detail. The critical heat flux are also predicted in vertical upward round tube according to the theory that the dryout in vertical annular flow emerges at the point where the film is depleted due to the integrating result of entrainment, deposition and evaporation. The influence of mass flux, inlet mass quality and tube diameter on critical heat flux is also predicted in this paper. Finally the predicted result of critical heat flux is compared with experimental data, and the theoretical CHF values are higher than that of experimental data, with error within 30%.

Devices to Mitigate Gas Entrainment in the Surge Tank of PFBR

Surge tank is one of the important components in the secondary circuit of a typical sodium cooled fast breeder reactor, provided to take care of pressure surges in case of a sodium water reaction in Steam Generators (SG). The blanket of argon cover gas at the top of the tank acts as a cushion for the surges. The argon gas above the free surface of sodium in the tank is a source of entrainment into the sodium which is undesirable from the consideration of effective heat transfer in Inter mediate Heat Exchanger (IHX) and SG cavitation in pumps and operational problems of continuous feed and bleed of cover gas, thus leading to unfavourable reactor operating conditions. To investigate the phenomenon of gas entrainment in surge tank, hydraulic experiments were conducted in water using 1/12 scaled model. The minimum height of liquid column in the tank when gas entrainment is completely avoided was established. Different methods to mitigate gas entrainment were tested in the model and a combination of porous plate and stiffener ring was found to be optimum in reducing the liquid column required to mitigate gas entrainment in surge tank.

Two-Dimensional Optical Measurement of Waves on Liquid Lithium Jet Simulating IFMIF Target Flow

Waves on a liquid-lithium jet flow, simulating a proposed high-energy beam target design, have been measured using an optical technique based on specular reflection of a single laser beam on the jet surface. The streamwise and spanwise fluctuations of the local free-surface slope were least-square fitted with a sinusoidal curve to makeup the signals lost due to the constriction in the optical arrangement. The waveform was estimated with an assumption that wave phase speed can be calculated using the dispersion relation for linear capillary-gravity waves. The direction of propagation on the jet surface was also evaluated so that the wave amplitudes, calculated by integral of slope angle signal, agree consistently in streamwise and spanwise direction. These measurements and analyses show that the waves at the measurement location for a jet velocity of 1.2 m/s can best be represented by oblique waves with an inclination of 1.23 rad, a wavelength of 3.8 mm and a wave amplitude of about 0.05 mm.