Band 86, Heft 4

A computational study has been carried out for predicting the behaviour of a pool fire source using the field-model based code Fire Dynamics Simulator (FDS). Time dependent velocity and temperature fields are predicted along with the resulting changes in the plume structure and its width. Firstly, a grid study was performed to find out the best grid size for this purpose. Then calculations were done which showed a very good agreement with earlier reported experimental based correlations for the temperature of the plume region. These studies have been extended to use this field-model based tools for modelling particular separate effect phenomena like puffing frequency and to validate against experimental data. There are several applications in nuclear industry like room fires, wildland fires, smoke or ash disposal, hydrogen transport in nuclear reactor containment, natural convection in building flows etc. In this paper the use of FDS with the advanced Large Eddy Simulation (LES) based CFD turbulence model is described for various applications: Fire simulation for Alpha storage, Bhabhatran teletherapy, pool fire for transport casks, fire PSA of a representative NPP, exhaust air fan buildings of a process plant and smoke dispersion in large fires around NPPs.

The characterization of oxidative degraded polystyrene-based resin (R1) and polyacrylic based resin (R2) resins in H 2 O 2 and HClO 4 degradation medium were made based on the kinetics and thermodynamic data obtained for the ion-isotopic exchange reactions using such resins. For the reactions performed by using resins degraded in H 2 O 2 medium, the reaction rate (k) values obtained for the fresh R1 (0.315 min –1 ) and R2 (0.187 min –1 ) resins decreases to 0.300 and 0.155 min –1 respectively for the resins degraded in 20% H 2 O 2 medium, which further decreases to 0.289 and 0.142 min –1 respectively for the resins degraded in 30% H 2 O 2 medium. A similar trend in the results were observed for the reactions performed by using the above resins degraded in HClO 4 medium. The higher values of k (min –1 ) and low values of various thermodynamic parameters for the ion-isotopic exchange reactions performed by using fresh and degraded polystyrene-based resin R1 resins suggests superior degradation stability as compared to polyacrylic based R2 resin.

Converting the weapon grade Plutonium from the U.S.A., Russia, U.K. etc. to M ixed OX ide fuel and using it in power reactors was seen as a feasible way to both dispose Plutonium and produce energy. Using Thorium-based fuels in CANDU has been investigated since early 1980’s, they were designed and tested in Canada as mixed ThO 2 -UO 2 (both LEU and HEU) and mixed ThO 2 -PuO 2 , (both reactor- and weapons-grade) ([1]). In this respect, Thorium might also be seen as a valuable driver for weapon grade Plutonium annihilation. Our goal was to investigate ThO 2 -PuO 2 MOX in the aim to propose a suitable fuel for the existing and future CANDU units in Romania. Both weapon grade and reactor grade Plutonium were considered as fissile drivers for Thorium. Since this is only an exploratory study, some key design parameters such as fuel pellet density and ThO 2 /PuO 2 ratio were considered to span over a certain range imposed by MOX fuel fabrication technology and limited Plutonium availability. Eighteen fuel compositions were considered and cell calculations were performed for 37 and 43-element bundles using several computer codes.

Higher order approximations of the Chebyshev polynomials of first kind (T N ) are used for the first time in calculation of the diffusion lengths of monoenergetic neutrons in a homogeneous slab. In the method, the diffusion lengths of the neutrons are calculated using various values of the c, the number of secondary neutrons per collision. First, the traditional Legendre polynomials (P N ) approximation and then the present T N method are used separately. The numerical results for the diffusion lengths are tabulated in the tables up to an order of N = 9. A brief comparison is also done between the results obtained from the present method and the ones in literature. The advantages of the present method can easily be observed from the good accordance between results given in the tables for comparison and its easily executable equations. For many of the c values, the results obtained from T N method are better than the results obtained from P N method.

Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MW th total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

This work investigates the capacity to detect safety levels in pressurized water reactors using neutronic and thermal-hydraulic calculations (PWR). The main neutronic parameters, such as criticality calculations, reactivity calculations, burn-up calculations, and power distribution calculations, have been calculated by using MCNPX based on the Monte Carlo method. The hot channel was determined from the power distribution. On the hot channel, the rod-centered sub-channel model was used to investigate the main thermal-hydraulic parameters such as the fuel, cladding, and coolant temperature distribution in both axial and radial directions, the coolant pressure drop, and the departure from nucleate boiling ratio (DNBR). Two techniques were used to solve the rod-centered sub-channel model. MATLAB code was used for the analytical methodology, and COMSOL-Multiphysics computer software was used for the numerical methodology. The numerical technique has demonstrated high accuracy in determining the reactor safety level consistent with the analytical approach.