Band 76, Heft 2

The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO 2 on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation.

In the present configuration of the calandria for the 700 MWe Kakrapara Nuclear Power Plant, moderator inlet diffusers are directed upwards and the outlet is from the bottom of the calandria. Moderator circulation patterns and temperature distribution needs to be predicted to ensure adequate cooling margin for all channels. This study consists of two steps: at first, an optimized calculation scheme is obtained by comparison of the predicted results with the experimental data and by evaluating the fluid flow and temperature distribution. Then, in the second step, the analysis for the real 700 MWe IPHWR moderator under normal operating conditions has been performed with the optimized scheme. The present paper describes the methodology used for predicting the circulation pattern and temperature distribution in the moderator during normal operation using CFD code CFD-ACE+. The matrix of the calandria tubes in the core region is simplified to a porous media in which the momentum resistance model is used for pressure loss. The buoyancy effects due to internal heating and jet momentum effects through inlet nozzles have been considered in the analysis. The results show that the maximum temperature observed in the calandria is within the design limits during normal operation.

Single phase natural circulation in a rectangular loop is simulated using the PHOENICS code, a general purpose Computational Fluid Dynamics (CFD) code. The rectangular loop, having different operating power levels, has been modeled with the help of the Multiple Block Fine Grid Embedment (MBFGE) technique. The Co-located Co-variant Method (CCM) is used to simulate this loop in PHOENICS. Extensive experimental and CFD studies have been conducted on single phase natural circulation in a rectangular loop. The paper presents the results of three-dimensional CFD analysis for the prediction of steady state behavior in a rectangular loop and its comparison with experimental data. The results of code prediction and readily available experimental data show good agreement.

In water cooled power reactors, significant quantities of hydrogen could be produced following a severe accident (loss-of-coolant-accident along with non availability of Emergency Core Cooling System) from the reaction between steam and zirconium at high fuel clad temperature. In order to prevent the containment and other safety relevant components from incurring serious damage caused by a detonation of the hydrogen/air-mixture generated during a severe accident in water cooled power reactors, passive autocatalytic recombiners (PAR) are used for hydrogen removal in an increasing number of French, German and Russian plants. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces generating steam and heat. Numerous tests and simulations have been conducted in the past to investigate passive autocatalytic recombiners behaviour in situations representative of severe accidents. Numerical models were developed from the experimental data for codes like COCOSYS or ASTEC in order to optimise the passive autocatalytic recombiners location and to assess the efficiency of passive autocatalytic recombiners implementation in different scenarios. However, these models are usually simple (black-box type) and based on the manufacturer's correlation to calculate the hydrogen depletion rate. Recently, uses of enhanced CFD models have shown significant improvements towards modeling such phenomenon in complex geometry. The work presents CFD analysis of interaction of a representative nuclear power plant containment atmosphere with passive autocatalytic recombiners simulated using the commercial Computational Fluid Dynamics code for PAR Interaction Studies (PARIS benchmarks) exercise. A two-dimensional geometrical model of the simulation domain was used. The containment was represented by an adiabatic rectangular box with two PAR located at intermediate elevations near opposite walls. The flow in the simulation domain was modelled as single-phase. The results of the simulations are presented and analysed.

In the thermal hydraulics codes developed for fire safety analysis and for containment thermal hydraulic analysis, junctions in the multi-compartment geometries is often modeled as uni-directional junctions. However, ceiling junctions are known to depict unstable/oscillatory bi-directional flow behavior. Detailed investigations have been carried out to understand the unstable flow behaviour of a junction by analyzing an earlier reported experiment and its subsequent two dimensional numerical RANS based study of fire in an enclosure. The authors attempt more realistic and desired three dimensional and inherently transient large eddy simulations using a computer code Fire Dynamics Simulator (FDS). The paper presents the details of the analysis, the results obtained and further studies required to be conducted so that the findings can be applied to the fire/containment thermal hydraulics analysis codes successfully.

The thermal pollution arising out of discharge of hot water from the power plant condensers into the natural water bodies such as rivers, lakes, reservoirs, oceans etc. has been a serious concern to environmentalists ever since the plants started operating world over. In the past forty to fifty years, the methods of calculations for predicting the velocity and temperature fields in the affected regions of the stagnant/flowing water bodies have undergone a significant improvement. Currently, use of Computational Fluid Dynamics (CFD) codes for performing these calculations is gaining popularity. However, several factors such as the assumed computational domain and its discretisation, the boundary conditions used, representation of hydrodynamic characteristics (laminar/turbulent, buoyant/non-buoyant), etc. have a strong influence on the accuracy of predictions by such a model. A CFD code STAR-CD has been used for analyzing the thermal plume behaviour in the Kadra reservoir at Kaiga Atomic Power Station (KAPS). The predictions from these calculations of two units in operation have been found to be in good agreement with the site data made available from earlier studies. The present paper briefly describes the model developed using STAR-CD and results obtained for the Kadra reservoir at KAPS.

Scalar dispersion in the atmosphere is an important area wherein different approaches are followed in development of good analytical models. The analyses based on Computational Fluid Dynamics (CFD) codes offer an opportunity of model development based on first principles of physics and hence such models have an edge over the existing models. Both forward and backward calculation methods are being developed for atmospheric dispersion around NPPs at BARC. Forward modeling methods, which describe the atmospheric transport from sources to receptors, use forward-running transport and dispersion models or computational fluid dynamics models which are run many times, and the resulting dispersion field is compared to observations from multiple sensors. Backward or inverse modeling methods use only one model run in the reverse direction from the receptors to estimate the upwind sources. Inverse modeling methods include adjoint and tangent linear models, Kalman filters, and variational data assimilation, and neural network. The present paper is aimed at developing a new approach where the identified specific signatures at receptor points form the basis for source estimation or inversions. This approach is expected to reduce the large transient data sets to reduced and meaningful data sets. In fact this reduces the inherently transient data set into a time independent mean data set. Forward computations were carried out with CFD code for various cases to generate a large set of data to train the Artificial Neural Network (ANN). Specific signature analysis was carried out to find the parameters of interest for ANN training like peak concentration, time to reach peak concentration and time to fall. The ANN was trained with data and source strength and locations were predicted from ANN. The inverse problem was performed using the ANN approach in long range atmospheric dispersion. An illustration of application of CFD code for atmospheric dispersion studies for a hypothetical case is also included in the paper.

Sipping tests have been performed for fuel elements of the TR-2 reactor at Çekmece Nuclear Research and Training Center (ÇNRTC), in order to find out which one failed in the core. A sipping assembly has been constructed and placed in the pool of the TR-2 reactor. The assembly identifies leaking fuel elements by collecting and measuring 137 Cs that leak out from the defective fuel elements. 31 fuel elements in the reactor have been tested for the clad integrity. The measured 137 Cs activity of the fuel element with an identification number S-104 is a 10247 Bq/(0.3 l). This value is approximately 234 times greater than the average of the other tested fuel elements in the reactor.

In this work the neutron multiplication in a source driven subcritical nuclear system is studied for a particular physical model of symmetrical slab geometry. On both sides of a central spallation source region, bare fissile blankets are placed. The neutronics of the system is represented in terms of two-group diffusion equations. We elaborate on the development and evaluation of the multiplication factors for source and fission neutrons (and). The results are compared with the conventional multiplication factor of the fissile system. In the present static context, is important only from the safety point of view, whereas the other two factors are figures of merit measuring the effectiveness of a source neutron to start a fission chain and that of any one fission neutron to maintain a fading chain which further contributes to source multiplication.

101m Rh is one of such radionuclides that has been considered as a potential candidate for targeted radiotherapeutic use, due to its nuclear decay and chemical properties. Electrodeposition of rhodium metal on a copper backing was performed in acidic sulphate. The target was bombarded with a current intensity of 120 μA (E p = 29 → 25 MeV) for 30 min (60 μAh). Radiochemical methods were investigated to optimize the production of no-carrier-added 101 Pd/ 101m Rh. The use of a cyclotron target with radiochemical processes (i.e. electrodeposition, electrodissolution and ion-exchange column chromatography) were carried out to produce this radionuclide in high-specific activity and radiochemical form suitable for radiopharmaceutical syntheses.

In this study, neutron incident reaction cross sections for some target nuclei such as 24 Mg, 27 Al, 28 Si, 56 Fe, and 63 Cu used in neutron activation analysis have been investigated. The new calculations on the excitation functions of 24 Mg(n, p) 24 Na, 27 Al(n, p) 27 Mg, 27 Al(n, α) 24 Na, 28 Si(n, p) 28 Al, 56 Fe(n, p) 56 Mn, and 63 Cu(n, 2n) 62 Cu reactions have been carried out for incident neutron energies up to 20 MeV. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve the new geometry dependent hybrid model and the full exciton model. Equilibrium effects are calculated according to the Weisskopf–Ewing model. In the present work, reaction cross-sections have been calculated by using empirical formulas developed for energies of 14–15 MeV. The calculated results are discussed and compared with the experimental data taken from the EXFOR database.

The critical slab problem is investigated in one-speed neutron transport theory for reflecting boundary conditions using second kind Chebyshev polynomials approximation. The forward-backward-isotropic scattering kernel is used in a uniform homogeneous slab. The critical slab thicknesses for various values of the collision parameter, forward and backward scattering and reflection coefficient are given in the tables together with the ones available in literature and they are in good agreement.