Volume 71, Issue 5-6

Nuclear waste actinides can be used as a booster fissile fuel material in form of mixed fuel with thorium in a CANDU reactor in order to assure the initial criticality at startup. Two different fuel compositions have been found useful to provide sufficient reactor criticality over a long operation period: 1) 95% thoria (ThO 2 )+5% minor actinides MAO 2 and 2) 90% ThO 2 +5% MAO 2 +5% UO 2 . The latter allows a higher degree of nuclear safeguarding through denaturing the new 233 U fuel with 238 U. The temporal variation of the criticality k ∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 10 years. The criticality starts by k ∞ >1.3 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2 nd year and remains close to k ∞ = ∼1.06 for ∼10 years. After the 2 nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn up could be achieved with the same fuel material (up to 200000 MW.D/MT), provided that the fuel rod claddings would be replaced periodically (after every 50000 or 100000 MW.D/MT). The reactor criticality can be maintained until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.

Hydrogen distribution in the containment during severe accidents is an important safety issue of nuclear power plants operation. The detailed knowledge of the containment thermal-hydraulics is necessary for reliable predictions of pressure, temperature and gas distribution in the containment of the NPP during postulated accidents. For this purposes, the study of the International Standard Problem (ISP 47) has been started in 1999 with the test facilities TOSQAN, MISTRA and ThAI. The paper presents the COCOSYS nodalisation of the TOSQAN test facility. The results are analysed and compared with the experimental values.

Prediction of the onset of the flow instability (OFI) in steady and transient sub-cooled flow boiling is an important consideration in the design and operation of nuclear reactors especially material test reactors (MTR). In this study, a predictive model of OFI in the MTR has been developed. The model is based on both heat balance of bubble generation and condensation and force balance for the detached bubbles at the onset of significant void (OSV). Only one coefficient involved in the proposed model is identified by experimental data that covers the wide range of MTR operating conditions. The model results are compared with results of some previous models where the present model gives the lower deviation from the experimental data. A correlation for the heat flux at OFI is also developed based on the present model. The developed correlation gives lower deviation from the experimental data than the well-known correlation of Whittle and Forgan. The model is also used to predict the OFI locus during transient where it shows good agreement with the short transient of Lee and Bankoff as well.

The angular distribution of the radiation intensity and the related constants C and D∗ are calculated for a new choice of c(x) having the form of the Morse potential in quantum mechanics c(x) = 1 – a – be −2 αx + de − αx . We use the modified Eddington method. The radiation intensity in this method is given in terms of the unknown even and odd functions of the space variable × and the direction cosine. The coefficients of these functions depend only on the space variables and satisfy second order differential equations. We solve the resulting second order differential equation using the Nikiforov-Uvarov method. The method provides exact analytical expressions and is not previously used to solve radiation problems. The numerical results are listed in a table for both the constants C and D∗ and the albedo and in the limiting cases are compared with the homogeneous values.

Half-space albedo, slab albedo and critical slab problems will be solved for extremely anisotropic scattering using the analytical expressions of the F N method with a different numerical approximation. The infinite medium Green's function for the anisotropic scattering case obtained in terms of the infinite medium Green's function for the isotropic scattering is used. The numerical results are compared with data available in the literature.

In this work, the H N approximation is used to solve the Milne problem for both isotropic and linearly anisotropic scattering. For different specular and diffuse reflection coefficients the extrapolated endpoint is calculated. The numerical results obtained are compared with results already existing in the literature. It is shown that they are in good agreement.

Since April 1998, the OECD Nuclear Energy Agency (NEA) is operating the International Common Cause Data Exchange (ICDE) project. The objectives of this exchange comprise, to collect and analyze CCF events, to generate insights into root causes and derive preventing approaches, to strengthen experience feedback and, last but not least, to facilitate quantification. For a large number of about 1000ICDE events the fractions of complete CCFs have been displayed as a function of the degree of redundancy. Complete CCFs are relevant to PSA as in these cases the respective function fails and in many cases significant contributions to core damage can be expected. From the point of view of quantitative CCF modeling, the theoretical theoretical approach to of the empirically available complete CCF fractions represents a major challenge. The large number of events have been collected in the ICDE project for nine different types of components in nuclear power plants from ten different countries. Obviously, these components are operated under different conditions and technical specifications. Therefore, an approach is chosen which is based on low information distributions of the model parameters to account for the large variation throughout the component populations. The Process Oriented Simulation Model (POS) has been selected to carry out an analysis of complete CCFs. The consistency between the theoretical approach – considering its simplicity – and the data observed is satisfactory.

The nuclear fuel of the operating RBMK-1500 type reactor at Ignalina Nuclear Power Plant is a potential source of accumulated activity which can enter the surrounding environment. Due to the fact, that the high radiotoxic nuclear fuel fission products 90 Sr and 137 Cs are most important radioactive pollutants, this study is concentrated on the quantitative and qualitative investigation of the cumulative behaviour of all strontium and caesium isotopes during burn-up and cooling time of the nuclear fuel. An investigation is performed on the basis of modelling results, which are important for the risk assessment of environmental contamination and for the interpretation of the results of direct field measurements.

The present work is carried out in order to study the closure flow inside the core chimney of the Egypt Second Research Reactor (ETRR-2). Based on the finite difference technique, a two dimensional model is developed to simulate the coolant flow inside the chimney. The model is verified by FEHT finite element program. Then a study of different closure flow values inside the chimney was made using the developed model where a flow map is plotted showing the stagnation depth for each closure flow. The flow map shows that for a closure flow greater than 0.16 m 3 /h, no active water ascends from the core to the pool through the chimney. The model results are analyzed and discussed.

An experimental method for the determination of the long-term activation of concrete shielding has been performed for two TRIGA research reactors in Ljubljana, Slovenia and in Vienna, Austria. The activity of concrete was studied on concrete samples taken directly from the external shielding of the reactor body with consecutive activation under specified parameters. The present work intends to help minimize radiation exposure for the reactor staff engaged in dismantling the concrete shield of a typical TRIGA Mark II research reactor.