Volume 73, Issue 4

The H N method is used to calculate the partial fluxes relevant to radiative transfer in a plane parallel medium for linearly anisotropic scattering with specularly and diffusely reflecting boundaries. The solutions are obtained for the non-conservative case. The numerical values of the heat flux functions at the boundaries of the medium are calculated for the two source functions. The numerical results and the convergence of the H N method are compared with earlier results.

The spectral Green's function (SGF) method is used to solve numerically the transport of neutrons in an infinite homogeneous cylindrical domain. In the method, the scattering of neutrons is assumed to be isotropic and the transport equation in cylindrical geometry is reduced to plane-like transport equation by using our reasonable approach. Then, the plane-like transport equation is solved by using the SGF method and the results are compared with those already obtained by the Diamond Difference (DD) scheme to check the accuracy of the results. The agreement of the results is good and it can be concluded that the SGF method is very effective for the numerical solution of the transport equation.

The critical slab problem has been studied in the one-speed neutron transport equation with isotropic scattering by using the first kind of Chebyshev Polynomials. The moment criticality solutions were obtained for the uniform finite slab using Mark and Marshak type vacuum boundary conditions. The results obtained by this approximation are presented in tables which also include the results obtained by the P N method for comparison.

The pseudo-slab problem by using the H N method, proper neutron flux definitions and physical boundary conditions is used to study spherical albedo. The selected numerical values can be compared with data available in the literature. The sphere radius can be calculated for the secondary neutron number and for the spherical albedo.

The one speed, time independent neutron transport equation for slab geometry with quadratic anisotropic scattering kernel is considered. Albedo, transmission factor and criticality thickness are calculated by the H N method. The numerical results obtained are listed for different selected parameters. It is shown that the method is concise and leads to fast converging numerical results.

In this work an analytical solution for the time-dependent S N transport equation in a slab using the double Laplace transform technique is reported. This procedure leads to an analytical solution for the angular flux in integral form. Numerical simulations are presented and compared with results in the literature.

In the present work, the WIMSD code routine PIJM is modified to compute deterministic Dancoff factors by the collision probability definition in general arrangements of partially absorbing fuel rods. Collision probabilities are calculated by an efficient integration scheme of the third-order Bickley functions, which considers each cell region separately. The effectiveness of the method is assessed by comparing grey Dancoff factors as calculated by PIJM, with those available in the literature by the Monte Carlo method, for the irregular geometry of the Canadian CANDU and CANFLEX assemblies. Dancoff factors at several different fuel pin positions are found in very good agreement with the literature results.

In this study, new calculations of the excitation functions of 56 Fe(p,2n) 55 Co, 58 Fe(p,2n) 57 Co, 60 Ni(p,n) 60 Cu, 62 Ni(p,2n) 61 Cu, 68 Zn(p,2n) 67 Ga, 72 Ge(p,n) 72 As, 82 Kr(p,2n) 81 Rb, 124 Xe(p,2n) 123 Cs, 112 Cd(p,2n) 111 In and 203 Tl(p,3n) 201 Pb reactions have been carried for incident proton energies up to 50 MeV. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve the newly evaluated geometry-dependent hybrid model, hybrid model and the cascade exciton model. Equilibrium effects are calculated according to the Weisskopf–Ewing model. The calculated results are compared with the experimental data taken from the literature.

The prediction of the time-dependent failure rate has been studied, taking into account the operational history of a component used in applications such as system modeling in a probabilistic safety analysis in order to evaluate the impact of equipment aging and maintenance strategies on the risk measures considered. We have selected a time-dependent model for the failure rate which is based on the Weibull distribution and the principles of proportional age reduction by equipment overhauls. Estimation of the parameters that determine the failure rate is considered, including the definition of the operational history model and likelihood function for the Bayesian analysis of parameters for normally operating repairable components. The operational history is provided as a time axis with defined times of overhauls and failures. An example for demonstration is described with prediction of the future behavior for seven different operational histories.

Processes occurring during the reactor operation and accidents change the thermal, chemical and other properties in nuclear fuel rods. This paper describes the processes in the nuclear fuel rods and tools for simulation at these processes. Fuel rods used at Ignalina NPP were simulated with FEMAXI-V (Japan) and TESPA-ROD (Germany) codes. Developed models were employed for the analysis of processes in fuel rods in case of large LOCA Beyond Design Basis Accident (BDBA). The developed models and results of the analysis are presented in this paper.

Heavy metal molten salts have a good potential to be used in fusion reactors due to their capability of increasing fusion power and producing fissile fuel. In this study, the effect of various molten salts bearing nuclear fuel (99.5 % Flibe + 0.5 % Weapon grade (WG) PuF 4 , 99 % Flibe + 1 % WG PuF 4 , 99.5 % Flibe + 0.5 % Reactor Grade (RG) PuF 4 , 99 % Flibe + 1 % RG PuF 4 , 75 % LiF + 25 %ThF 4 , 75 % LiF + 24 % ThF 4 + 1 % 233 UF 4 and 75 % LiF + 23 %ThF 4 + 2 % 233 UF 4 ) on the neutronic performance of a magnetic fusion reactor called ARIES-RS was investigated. Neutron transport calculations were performed with the help of the code, Scale4.3. Numerical results showed that a significant enhancement on the neutronic performance was achieved with respect to the energy multiplication and fissile fuel breeding by keeping the sufficient tritium for the (DT) driver compared to the pure ARIES-RS fusion reactor for all investigated coolants except for the salts of 75 % LiF + 25 % ThF 4 and 75 % LiF + 24 % ThF 4 + 1 % 233 UF 4 . Since these two salts did not satisfy tritium self-sufficiency.