Band 74, Heft 5-6

The European Platform for NUclear REactor SIMulations, (NURESIM project 2005 – 2008) addressed the creation of a Common European Standard Software Platform for modelling, recording, and recovering computer data for nuclear reactors simulations. One work package of the project was dedicated to the analysis and improvement of CFD capabilities for the simulation of two-phase PTS problems. Some SB-LOCA scenarios lead to a situation in which the cold leg is partially or totally uncovered when the Emergency Core Cooling injection is activated. The resulting complex two phase flow can be divided in characteristic flow regions: the jet flow with a free surface between steam and water, the zone of jet impingement, the horizontal two-phase flow and the flow in the downcomer. Many phenomena have to be reflected in a simulation of each separate region, but also when the simulations are coupled reflecting the integral process which is required to predict the thermal loads at the RPV wall. After analyzing the experimental database available for CFD model development and validation and identifying shortcomings of the models different activities were dedicated to the simulation of single flow regions as well as the integral flow. Based on these experiences recommendations for the CFD-simulation of the two-phase PTS problem were obtained.

This article focuses on the validation of the Lumped Parameter (LP) computer codes COCOSYS and ASTEC in the field of containment phenomena during severe accidents. The release of hydrogen (H 2 ) as an open jet from the primary circuit into the containment might cause atmospheric H 2 -stratifications which in turn may result in an ignitable atmospheric mixture. If this mixture is ignited, the integrity of the containment can be endangered due to the pressure and temperature rise, resulting from the combustion process. The emphasis of the paper lies on these three phenomena referring to possible successive phases of a loss of coolant accident.

The investigation of insulation debris generation, transport and sedimentation becomes important with regard to reactor safety research for PWR and BWR, when considering the long-term behaviour of emergency core cooling systems during all types of loss of coolant accidents. A joint research project on such questions is being performed in cooperation between the University of Applied Sciences Zittau/Görlitz and the Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation of particle transport phenomena in coolant flow and the development of CFD models for its description. While the experiments are performed at the University at Zittau/Görlitz, the theoretical modelling efforts are concentrated at Forschungszentrum Dresden-Rossendorf. In the current paper the basic concepts for CFD modelling are described and feasibility studies are presented.

Detailed CFD model for the head parts of the VVER-440 fuel assemblies with burnable poison has been developed. The coolant mixing was analyzed in some typical assemblies with this model and the signals of the in-core thermocouples above the selected assemblies were calculated. The investigations pointed out that the mixing is intensive in these assembly heads but the coolant is not perfectly mixed before reaching the thermocouples. Significant differences between the outlet average coolant temperatures and the thermocouple signals were revealed in the case of the fresh fuels. These deviations can cause about 6 % underestimations in the online monitored assembly powers unless a proper correction is introduced. The coolant mixing was also studied by means of numerical tracers and weight factors of selected rod bundle regions for the in-core thermocouple were determined. Using these weight factors and the outlet enthalpies of the assemblies' subchannels, the thermocouple signals can be corrected.

The prediction of time dependent standby failure rates was studied, taking into account the operational history of a component. These studies are important for applications such as system modeling in probabilistic safety analysis to evaluate the impact of equipment aging and maintenance strategies on the risk measures (e. g. reactor core damage frequency) considered. The time dependent model for the standby failure rate is defined based on the Weibull distribution and the principles of proportional age reduction by equipment overhauls. The parameters which determine the standby failure rate are estimated, including the definition of the operational history model and likelihood function for Bayesian analysis of parameters for periodically tested standby repairable components. The operational history is provided as time axis with defined times of overhauls, surveillance tests and failures. Assessment of time dependent unavailability due to the failure of periodically tested standby components is described. As an example, the prediction of the future behavior of components for seven different operational histories is described.

A closed loop auto control system has been developed and implemented at the Pakistan Research Reactor-2 (PARR-2). For interfacing of the site signals with a computer, a data acquisition card was employed. The system utilizes nine signals and, on their basis, carries out continuous on-line analysis, maintaining safe operation. On-line reactivity measurement is included in this software. Also, it generates control signals for keeping the reactor at the desired power level in auto-mode round-the-clock without human intervention. In case of abnormal conditions for either power or temperature or radiation level, alarms are initiated and if their levels reach beyond prescribed safe operation regime, automatic shutdown sequence is initiated. The control system has been thoroughly tested for various postulated scenarios. The test results have been in good agreement with the expected response.

Resolving hydrogen related safety issues, pertaining to nuclear reactor safety have been an important area of research world over for the past decade. Studies on hydrogen transport behavior and development of hydrogen mitigation systems are still being pursued actively in various research labs, including Bhabha Atomic Research Centre (BARC), in India. The Passive Catalytic Recombiner is one of such hydrogen mitigating devices consisting of catalyst surfaces arranged in an open-ended enclosure. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces and the heat of the reaction produces a natural convection flow through the enclosure. The present study aims for performing a CFD simulation to obtain the temperature distribution inside the recombiner. A 3D CFD model has been developed to study the mechanism of catalytic recombination and has been tested for literature quoted experiments. A parametric study has been performed for a particular recombiner geometry for various inlet conditions. Salient features of the simplified CFD model developed at BARC and results of the present model calculations are presented in this paper.

Using the method of “affine solutions” three interesting phenomena in a CANDLE Burn-Up strategy were studied and were cast analytically into mathematical formlulae: a) the problem of two burn-up waves one behind the other (“enslavement principle”) b) the problem of the frontal collision of two burn-up waves (“Scattering”), and c) the problem of two identical burn-wave twins moving apart from each other after central ignition (“Breathing Out Phenomenon”). To the knowledge of the author this is the first time where such affine problems of the CANDLE Burn-Up wave physics were solved.

The radiological consequences of operational atmospheric discharges from nuclear power plants are analyzed by using coupled long-range atmospheric dispersion models with time-dependent dose and risk assessment models. The coupled model is applied to atmospheric discharges from twelve assumed nuclear power plants located at three different sites in Turkey. Individual as well as collective doses and risks are analyzed. The effects of multi-location food supply and interregional movements of people in the computational domain are also investigated while computing individual effective dose and risk values. The values computed are found to be negligible compared with values from natural background and exposures from nuclear weapon testing.

In this study, Chebyshev polynomial method has been applied in the acceleration of the outer iterations for the numerical solution of the two dimensional neutron diffusion equation by the boundary element method. Boundary element discretization of the diffusion equation results in a full and nonsymmetric coefficient matrix which is quite different than the sparse and symmetrical matrices of the finite element and finite difference methods to which Chebyshev acceleration has been classically applied. To assess the merit of the Chebyshev method in the case of boundary element discretization of the multigroup diffusion equation constitutes the main objective of the study. Numerical experimentation has established that the Chebyshev acceleration is effective also in case of boundary element discretization despite the dissimilarity of the coefficient matrices compared to the more conventional methods.

Series expansion of the infinite medium Green's function in terms of Legendre polynomials is used to solve the half-space albedo problem. The importance of this procedure is that it can be used for the solution of many different physical problems in plane geometry and can be extended to find the solution of the problems in other geometries.

In order to develop a general procedure, the scattering kernel is described in a consistent manner. By using this scattering kernel, eigenvalues and eigenfunctions and critical slab thicknesses are computed with a conventional spherical harmonics method (P N ). The basis of this method relates to the generating function of Legendre polynomials. This article has improved analytical methods and numerical values. Numerical results indicate that the critical thickness varies with forward isotropic scattering. Results are discussed and compared with isotropic and anisotropic results using various methods available in the literature. Eigenvalues and critical thicknesses are calculated using the new phase function instead of conventional expansions.

This paper addresses the training and retraining programme for research reactor operating personnel. The factors that need to be considered in establishing an effective organizational structure for research reactor operation are discussed as they form the basis for the development of the initial training and retraining programme for the operating personnel. Different elements for a successful implementation of this programme such as training system, materials, and methods are also discussed. Application of these elements to the case of the Egypt Second Research Reactor (ETRR-2), which is a relatively large research reactor, is illustrated. The detailed content of the training programme is presented along with discussions that include the applicability of this programme to smaller research reactors.