Volume 79, Issue 6

Evaluated proton, deuteron, triton, 3 He and 4 He production cross-section data were prepared for natural titanium irradiated with protons at the energy up to 3 GeV. The evaluated data were obtained from the analysis of results of nuclear model calculations, available measured, and systematic data. The calculations of gas production cross-sections were carried out using the intranuclear cascade model (INC), the pre-equilibrium exciton model, geometry dependence hybrid (GDH) model, Weisskopf-Ewing (WE) model, and the Hauser-Feshbach model. CASCADE, TALYS, ALICE/ASH, and ALICE/HMS codes were used for calculations. Available experimental and systematic data, model calculations were used to evaluate gas production cross-section. The evaluation was carried out using statistical methods implemented in the BEKED code package developed at KIT/INR. The evaluated cross-sections were compared with the data from evaluated data libraries.

In this study, the fluids were composed with increased mole fractions of a mixture of molten salt: heavy metals 99 – 95 % Li 20 Sn 80 −1 – 5 % SFG-Pu, 99 – 95 % Li 20 Sn 80 −1 – 5 % SFG-PuF 4 , 99 – 95 % Li 20 Sn 80 −1 – 5 % SFG-PuO 2 . In this study, the effect on conversion of each isotope ( 238–242 Pu) in spent fuel grade plutonium by (n,γ) reactions was investigated in liquid first wall, blanket and shield zones of the designed hybrid reactor system. Beryllium (Be) is the neutron multiplier by (n,2n) reactions. The Be zone used was 3 cm thick. 9Cr2WVT, a ferritic steel, is used as structural material. Three-dimensional nucleonic calculations were performed by using the most recent versions of the MCNPX-2.7.0 Monte Carlo code and the nuclear data library ENDF/B-VII.0.

This paper aims to investigate 232 Th/ 233 U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. 232 Th/ 235 U/ 238 U oxide mixture was considered as fuel in the core, when the mass fraction of 232 Th was increased as 0.05–0.1–0.2–0.3–0.4 respectively, the mass fraction of 238 U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the 232 Th, 233 U, 238 U, 237 Np, 239 Pu, 241 Am and 244 Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

Due to limitation of computation resources and/or computation time many thermal problems require to use simplified geometrical models with equivalent thermal properties. A new method for definition of equivalent thermal conductivity of spent nuclear fuel storage casks is proposed. It is based on solving the inverse heat conduction problem. For the proposed method two approaches for equivalent thermal conductivity definition were considered. In the first approach a simplified model in conjugate formulation is used, in the second approach a simplified model of solid body which allows an analytical solution is used. For safety ensuring during all time of spent nuclear fuel storage the equivalent thermal conductivity was calculated for different storage years. The calculated equivalent thermal conductivities can be used in thermal researches for dry spent nuclear fuel storage safety.

Recently, many formalizations have been developed to explain some reaction mechanisms such as neutron or proton induced reactions. In this study, (d, n) reaction mechanism was investigated developing the cross-section systematic formula. The (d, n) cross sections at 8.6 MeV deuteron energy have been calculated using these formula for some target nuclei (A = 10∼123). The cross sections calculated are important for the development of fusion reactor technology. Obtained calculation results from the formula have been also compared and discussed with results calculated by the codes ALICE/ASH (equilibrium and pre-equilibrium theories) and TENDL-2012 (Talys-based), and with results of available experimental values. It is shown, that the suggested empirical equation at the 8.6 MeV deuteron energy works very well for a quick estimation of (d, n) nuclear cross sections.

In the present work we solve in analytical representation the three dimensional neutron kinetic diffusion problem in rectangular Cartesian geometry for homogeneous and bounded domains for any number of energy groups and precursor concentrations. The solution in analytical representation is constructed using a hierarchical procedure, i. e. the original problem is reduced to a problem previously solved by the authors making use of a combination of the spectral method and a recursive decomposition approach. Time dependent absorption cross sections of the thermal energy group are considered with step, ramp and Chebyshev polynomial variations. For these three cases, we present numerical results and discuss convergence properties and compare our results to those available in the literature.

The thermoluminescence response of Dysprosium doped hydroxyapatite samples with different mol percentage of 0.5, 1 and 2 were studied and compared with the pure hydroxyapatite. The samples were objected to 60 Co gamma rays irradiation with doses of 100 mGy to 10 Gy. The main peak in the sample glow curves were obtained at 310 °C. The sensitivity of the 1 mol% Dy doped hydroxyapatite samples show the highest thermoluminescence response. Fading behavior of the irradiated samples was also studied. The experimental results show that the synthetic Dy-doped hydroxyapatite obtained by the hydrolysis method may be used in gamma radiation dosimetry.

Radioanalytical techniques using 131 I and 82 Br as tracer isotopes were applied to study the kinetics of iodide and bromide ion-isotopic exchange reactions taking place between the external labeled ionic solution and the resin surface. The results indicate low values of specific reaction rate (min −1 ), amount of ion exchanged (mmol) and initial rate of ion exchange (mmol/min) for bromide ion-isotopic exchange reaction as compared to that obtained for iodide ion-isotopic exchange reaction. It was observed that for iodide ion-isotopic exchange reaction performed at 35.0 °C, 1 000 g of ion exchange resins and 0.002 mol/L labeled iodide ion solution, the values of specific reaction rate (min −1 ), amount of iodide ion exchanged (mmol), initial rate of iodide ion exchange (mmol/min) and log K d were 0.340, 0.394, 0.134 and 20.2 respectively for Purolite NRW-5050 resin, which was higher than the respective values of 0.216, 0.290, 0.063 and 18.2 as that obtained by using Duolite A-611. The results of present investigation indicate that during the two ion-isotopic exchange reactions, for both the resins, there exists a strong positive linear correlation between amount of ions exchanged and concentration of ionic solution; and strong negative correlation between amount of ions exchanged and temperature of exchanging medium. From the results it appears that as compared to Duolite A-611 resins, Purolite NRW-5050 resins shows superior performance under identical experimental conditions.

This paper presents a new approach based on a binary mixed integer coded genetic algorithm in conjunction with the weighted sum method for multi-objective optimization of fuel loading patterns for nuclear research reactors. The proposed genetic algorithm works with two types of chromosomes: binary and integer chromosomes, and consists of two types of genetic operators: one working on binary chromosomes and the other working on integer chromosomes. The algorithm automatically searches for the most suitable weighting factors of the weighting function and the optimal fuel loading patterns in the search process. Illustrative calculations are implemented for a research reactor type TRIGA MARK II loaded with the Russian VVR-M2 fuels. Results show that the proposed genetic algorithm can successfully search for both the best weighting factors and a set of approximate optimal loading patterns that maximize the effective multiplication factor and minimize the power peaking factor while satisfying operational and safety constraints for the research reactor.

This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

In this research, the simulation of one-sixth of the VVER-1000 reactor core is carried out by the WIMS-D4 nuclear code, based on the symmetry of the core and information obtained from the Final Safety Analysis Report. The atomic densities of some important nuclear materials and fission poisons are calculated by the WIMS-D4 code at the end of the first fuel cycle. In addition, the cross sections of some nuclides are obtained by WIMS-D4, and they are transferred into the CITATION code as inputs. In the next stage, neutron flux and reactor power are calculated by the CITATION code in Cold Zero Power and Hot Full Power status, and subsequently the heat flux of the core is obtained. Then the results are returned again into the extended program cycle. Finally, the heat flux of the core is inputted into the COBRA code, and the temperatures of fuel, clad and coolant are calculated along the various distances applying the COBRA thermal hydraulic code, through the results of the CITATION code and also initial data as default created from the Final Safety Analysis Report. In conclusion, there are some interesting outcomes resulting form the obtained results.