Verification of the WECHSL code on melt-concrete interaction and application to the core melt accident

The WECHSL code developed at the Kernforschungszentrum Karlsruhe describes the interaction of a hot melt with concrete structures and serves for upgrading from the limited experimental scale to reactor dimensions. After presenting the principles of concrete decomposition and of the heat transfer between a molten pool and a concrete structure, the modifications of the WECHSL code, which proved to be necessary during the verification process by means of the BETA experimental program, are outlined. Post test calculations in comparison with the experimental results are given both for an experiment with high power input, where completely liquefied melt layers interact with the concrete crucible, and for an experiment with low power input, where the heat transfer is governed by crust formation. Finally, computed results for the low pressure path of a core melt accident occurring in a German 1300 MWe standard PWR are given. Moreover, questions concerning a possible penetration of the reactor basemat are discussed under special consideration of the Chernobyl accident.     

Gas dynamics models for an oscillating gaseous core fission reactor

Two one-dimensional models are developed for the investigation of the gas dynamical behaviour of the fuel gas in a cylindrical gaseous core fission reactor. By numerical and analytical calculations, it is shown that, for the case that a direct energy extraction mechanisn (such as MHD) is not present, increasing density oscillations occur in the gas. Also an estimation is made of the attainable direct energy conversion efficiency, for the case that a direct energy extraction mechanism is present.     

BETA experiments on melt-concrete interaction: the role of zirconium and the potential sump water contact during core melt-down accidents

In the BETA test facility of Kernforschungszentrum Karlsruhe, prototypical core melts can be simulated in concrete structures sufficient in size to allow a computer-code-assisted extrapolation to be made to the reactor geometry.Three experiments have been carried out to investigate special aspects of molten corium interacting with concrete. The investigations and measurements show the dominance of Zr oxidation during concrete attack by the chemical reduction of SiO₂ to elemental Si and the subsequent Si oxidation by the gases from the concrete.Additionally, the failure of a cylindrical concrete wall was studied, which is eroded on the inner side by a heated melt while being cooled outside by stagnant water. In the experiment wall failure occurs and the melt relocates into the water annulus.Application of the experimental results to light-water reactor severe accidents is discussed.     

Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling).     

On the dynamics of unstable film boiling

A recently developed feedback system for the control of boiling heat transfer made possible the experiental investigation of process dynamics as well as static characteristics in all boiling regions.The region in which the process is normally unstable has now been studied experimentally for water boiling on a wire. A process transfer function previously developed analytically for this region gives quite good agreement between analytical and experimental transfer functions when heater end effects are taken into account.     

On the dynamics of film region flow boiling

A transfer function to describe boiling dynamics has been developed analytically. A system for control of the heater temperature in a boiling process was developed and constructed. Using this system, dynamic measurements were performed for pool boiling and experimental transfer functions were developed. Flow boiling dynamics were then investigated. Results obtained for nucleate and negative h region operation have been presented. The present study concerns film region flow boiling dynamics.     

Gas entrainment at the free surface of a liquid: entrainment inception at a vortex with an unstable gas core

Experiments have been performed using sodium, water, white spirit and Freon 113 in a simple small-scale system, to determine the conditions necessary for gas entrainment inception at a vortex with an unstable gas core. The results demonstrate that surface tension effects play a major role and hence they must not be ignored when developing an entrainment-free system.     

Seismic model test of the GCFR core and core support structure

A 15% scale model was constructed to study the dynamic structural behavior of the GCFR (gas cooled fast breeder reactor) core support structure during seismic excitation. The model contains a perforated aluminum plate with a diameter of 20 in. and 265 model core elements constructed from 7/8 in.-diameter aluminum tubes. The proper frequency and mass ratios of the core elements and the perforated plate was ensured by placing steel inserts in the tubes. The natural frequencies, mode shapes and damping factors were individually measured for each of the components and for the complete system. Harmonic and simplified seismic forcing functions were applied to study the dynamic behavior of the core and its support structure. The test results were compared with both analytical and computer code results. Applying thick plate theory, the effective elastic modulus is 27% lower than that given in the ASME code. The resonant frequencies and the mode shapes of the “combined” core and grid plate assembly were also calculated. Applying thick plate theory to the analytical method, the two lowest frequencies were determined and the comparison with the test results shows differences od 3 and 6%.     

Equivalent circuit model for the insulated core transformer

The design of the insulated core transformer(ICT)needs to consider the flux leakage effects.An equivalent linear circuit model is proposed based on the principle of duality.It is composed by two types of leakage inductances:conventional leakage between windings and special leakage introduced mainly by the insulation gaps.The values of leakage inductances depend on the dimensions of the core,gaps,or windings and the property of magnetic materials.The circuit allows for quantitatively evaluating influences of ICT internal parameters on its output properties.The winding self- and mutual inductance matrix is mathematically converted to derive the inductance formula.As an example,the leakage parameters of a sixstage two-dimensional(2D) ICT are calculated and analyzed.     

Numerical solution of the Yamanouchi model for core spray cooling

The Yamanouchi model is used in the evaluation of the emergency spray system for a postulated loss-of-coolant accident for boiling water reactors. As originally proposed, Yamanouchi developed his model to predict the progressive rewetting of fuel rods exposed to spray cooling. The model has been adapted to the rewetting of the metal wall shroud which encloses the fuel rod bundle. Yamanouchi's analytical solution is valid only after a constant velocity front of the wetted surface is attained. This note shows that the time necessary to establish the constant velocity front is not significant.     

高温气冷堆热气导管的结构稳定性分析

热气导管是高温气冷堆中氦气循环的重要流道,其在各工况下的结构完整性与稳定性关系到反应堆是否能运行安全。本文详细分析了热气导管在事故工况下所承受的压力载荷,包括绝热纤维对管壁的压力以及一回路失压事故时发生氦气泄漏产生压力;并根据得到的压力载荷计算了热气导管承受外压时的结构稳定性。计算结果表明热气导管在事故工况的压力载荷作用下能够保持结构的完整性并且不会发生失稳。     

Radiation effects at the HfO2-MgO interface

A preliminary investigation into radiation effects at the HfO₂-MgO interface, using classical molecular dynamics, is described. This composite system is representative of a dispersion nuclear fuel form concept being investigated for its potential in easing separations and reprocessing. During experiments involving ion bombardment of the interface a section of the top layer of HfO₂ has been seen to break away, resulting in a large (about 10 μm) crater with a raised central region. Computer simulations, using molecular dynamics, have been carried out on three separate models in an attempt to understand this behaviour. The first model investigates single atom bombardment of the interface. Cascades involving clusters of atoms are also investigated and finally a model of delamination at the interface is considered.     

A simplified model for the dynamics of a BWR

A model has been developed to derive the dynamic characteristics of a BWR with natural circulation. The model is based on the basic physical processes that govern reactor dynamics. The actual values for the model parameters are estimated from experimental and theoretical data. The model enables the computation of transfer functions of reactivity and steam flow to power and pressure. The sensitivity of these transfer functions to changes in model parameters is discussed.     

Influence of the nearest environment of the core on the power pulse dynamics in the IBR-2 reactor

An analysis of the IBR-2 reactor power pulse shape measured over the entire dynamic range of neutron flux variation (10⁴), i.e. from the maximum pulse power to the background power between pulses, has been carried out. Three variants of the model describing the reactor dynamics during the power pulse have been investigated. The best approximation to the experimental data has been obtained by adding to six equations describing the effect of delayed neutrons on the power pulse of two analogous ones describing the effect of the neutrons reflected from the structural elements of the reactor. It is shown that the most probable source of additional groups of neutrons may be the neutron moderators enveloping the core as well as the elements of the biological concrete shielding that are the closest to the core. These additional groups of neutrons influence essentially the formation of the power pulse.     

Seismic model of the gas cooled fast breeder reactor core support structure

The modelling technique for the seismic analysis of the core support structure of a gas-cooled fast breeder reactor is developed. The core support structure consists of the support cylinder and a perforated grid plate to which 265 fuel and blanket elements are clamped as cantilevered beams. The analysis of the core support structure consists of three steps: (1) analysis of the grid plate, (2) analysis of the core elements, and (3) modal synthesis.The first step in developing a solution to the problem is to assume that the core elements (fuel and blanket) are attached to the grid plate as rigid rods. In this case the influence of the rigid rods can be represented by their masses and rotary inertias. The solution of this problem was developed by applying the dynamic theory of grid plates. This was accomplished by generalizing the Reissner-Mindlin thick-plate theory with orthotropic constants and then modifying the formulations of the rotary inertia expressions to include the rotary inertia effects of the core elements. The numerical results showed that the grid plate's fundamental frequency is in the range of the fundamental frequencies of the core elements so that a dynamic coupling effect exists. Because of this dynamic coupling effect the elastic properties of the grid plate must be included in the seismic analysis of the GCFR'The second step was to develop a mathematical model of the grid-plate core-element system using the method of Rayleigh-Ritz. In this model the elastic coupling effect of the core elements was included.For the final application of the theory, the exact solution of the elastic plate with rigid rods was simulated on the computer by applying the elastic rotary inertias of the core elements to the model of the grid plate. With this technique it is possible to model the grid plate with a reasonable number of fuel and blanket elements and to replace the missing core elements with their equivalent effective rotary inertias. The method includes the capability of modeling the different mass, damping and elastic properties of the fuel and blanket elements.Comparing the results of the present analysis with the preliminary simple spring-mass core model, the amplitudes of vibration obtained, in some cases in the present analysis, are a factor of ten smaller than was previously computed. Applying this more elaborate analysis will lead to a simpler and less expensive design.     

A simple mathematical model of the core criticality control by burnable poisons

A simple mathematical model is proposed and developed for the core criticality control by burnable poisons (BPs) distributed only throughout the peripheral region of the core while its central region remains free from BPs. The numerical burnup calculations confirm the effectiveness of the considered BP distribution for the criticality control of nuclear reactors.     

Development of the critical film thickness correlation for an advanced annular film mechanistic dryout model applicable to MARS code

It is generally assumed in the mechanistic film dryout model that the critical heat flux (CHF) arises when liquid film calculated from evaporation, droplet entrainment and deposition gets dryout. The dryout of film is usually assumed when film thickness becomes zero. However, it was indicated that the complete dryout assumption can estimate CHF well for uniform heating case but cannot simulate accurately for non-uniform heating case. The critical film thickness concept may be an appropriate approach physically because there is a possibility of instantaneous disappearance of liquid film when it gets very thin. Therefore, a critical dryout film thickness correlation was developed to properly model dryout phenomenon together with MARS code based on experimental data. The modified version of MARS implementing a newly developed critical dryout film thickness correlation was assessed using various dryout data including those of non-uniform heating case and flow reduction transient test. The prediction results showed improved agreement with the experimental data.     

Theoretical analysis of the fluid mixing at the core inlet in rolling motion

The fluid mixing at the reactor core in rolling motion and steady state is investigated numerically with CFX12.0. The CFD results are validated with experimental data in steady state. In rolling motion, the fluid mixing factor at the center of the core oscillates in a cosine function, but the variation of the fluid mixing factor surrounding the core is not regular. The variation amplitude of the fluid mixing factor next to the boundary line of fluid mixing is the most significant. The variation of fluid mixing factor increases with the increasing of additional force. The increasing of Reynolds number could depress the effect of rolling motion on the fluid mixing.     

Simulation tests for temperature mixing in a core bottom model of the HTR-Module

Interatom and Siemens are developing a helium-cooled Modular High Temperature Reactor. Under nominal operating conditions temperature differences of up to 120°C will occur in the 700°C hot helium flow leaving the core. In addition, cold gas leakages into the hot gas header can produce even higher temperature differences in the coolant flow. At the outlet of the reactor only a very low temperature difference of maximum ±15°C is allowed in order to avoid damages at the heat exchanging components due to alternating thermal loads. Since it is not possible to calculate the complex flow behaviour, experimental investigations of the temperature mixing in the core bottom had to be carried out in order to guarantee the necessary reduction of temperature differences in the helium. The presented air simulation tests in a 1:2.9 scaled plexiglass model of the core bottom showed an extremely high mixing rate of the hot gas header and the hot gas duct of the reactor. The temperature mixing of the simulated coolant flow as well as the leakage flows was larger than 95%. Transfered to reactor conditions this means a temperature difference of only ±3°C for the main flow at a quite reasonable pressure drop. For the cold gas leakages temperature differences in the hot gas up to 400°C proved to be permissible. The results of the simulation experiments in the Aerodynamic Test Facility of Interatom permitted to design a shorter bottom reflector of the core.