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The P N method, the post-processed P N ( ppP N ) method, the discrete-ordinates ( S N ) method and the Monte Carlo method are used to solve the problem of the exterior of a sphere subject to a known incoming angular flux at the surface. This geometrical configuration introduces shadowing effects in the solution that show up as position-dependent angular-flux discontinuities in the angular variable. A brief description of the methods is given, including a prescription on how to normalize the angular flux, the scalar flux and the current in the S N and Monte Carlo codes used (ANISN and MCNP6), in order to allow for a proper comparison with the results from the other methods. Numerical results reported in graphical form for two test cases demonstrate that only the ppP N and Monte Carlo methods are able to reproduce well the profiles of the angular flux and the magnitudes of the angular-flux discontinuities. In addition, numerical results for the scalar flux and current are provided in tabular form for the test cases studied.

The heat transfer mechanism of natural convection has been extensively studied as a passive heat removal system of new nuclear power plants. Considering this aspect, the main objective of this study is to present an assessment of RELAP5 linear-equation solver under a transient two-fluid model for a two-phase natural circulation loop (NCL). For this assessment, three different approaches of linear-equation solvers for the hydrodynamic model are presented: the sparse matrix solver based on the Lower-Upper (LU) decomposition, the Border-Profile Lower Upper (BPLU) solver and the iterative method named Generalized Minimal Residual Method (GMRES). For comparison purposes, an experimental natural circulation loop made of glass tubes and using water as working fluid is analyzed. The onset of nucleate boiling observed during the experiment was predicted by all RELAP5 solvers as well as the representation of flow oscillations along the loop. Furthermore, it was noticed that the choice of the solver algorithm has a strong influence on the prediction of the two-phase natural circulation phenomena, since different wavelengths and amplitudes of flow instabilities were obtained for each approach.

The present study concerns a detailed analysis of flow boiling phenomena under high pressure systems using a two-fluid Eulerian approach provided by a Computational Fluid Dynamics (CFD) solver. For this purpose, a vertical heated pipe made of stainless steel with an internal diameter of 15.4 mm was considered as the modeled domain. Two different uniform heat fluxes and three saturation pressures were applied to the channel wall, whereas water mass flux of 900 kg/m2 s was considered for all simulation cases. The model was validated against a set of experimental data and results have indicated a promising use of the CFD technique for estimation of the wall temperature, the liquid bulk temperature and the location of the departure of nucleate boiling. Changes in factors applied in the modeling of the interfacial heat transfer coefficient and bubble departure frequency were suggested, allowing a better prediction of the void fraction along the heated channel. The commercial CFD solver FLUENT 14.5 was used for the model implementation.

When a certain void fraction value is reached in the two-phase flow regime, a problem occurs in the COBRA-EN code. This problem was observed in the drift-flux model option and interrupts code execution. Two solutions are proposed to solve the problem.

In this note, the neutron group fluxes are calculated through the implementation of the matrix form transport equation solution by the P N method and the main objective is to obtain accurate numerical results. Some results for two sample problems are graphically presented. A neutron flux to dose conversion is also proposed to simulate irradiation planning.

It is not possible to generate correct multigroup nuclear data with the GROUPR module of the NJOY system version 99.90 when the Maxwellian + 1/E + fission-spectrum weighting function is specified. It was detected that the calculation stops in the energy group that contains the break point between 1/E weighting and fission-spectrum weighting. A possible solution to bypass the error is proposed.

The computer codes presently utilized for spectrum average cross section calculations are discussed. A cost/benefit analysis of the NJOY system, the PREPRO2000 codes and the INTER code is presented. Considering the criteria of cost, availability, versatility and completeness the PREPRO2000 codes are considered the best option for researchers in developing countries involved with spectrum averaged cross section calculations.

The spherical-harmonics method is used to solve a class of multigroup criticality ( k -eigenvalue) problems in multislab geometry. The model includes scattering anisotropy of arbitrary order and allows reflective or vacuum outer boundaries. Numerical results of benchmark quality are reported for three sample problems that have been defined and used by other authors to study various transport methods for criticality calculations. A comparison with our results indicates that some of the earlier results are in error.

It was verified after a fuel burnup calculation with the WIMSD-5b code using the IAEA.LIB library that the computed average number of neutrons released per fission of Pu94242 shows up as a Not-a-Number (NaN) for some energy groups. As this problem does not permit the use of the generated multigroup microscopic cross sections by a reactor calculation code, the value of 1.0E−38 barns was attributed to all energy groups of the IAEA.LIB library that have null values of multigroup microscopic fission cross sections for this material.

Some problems encountered when processing material Th 90 232 from ENDF/B-VII.0 with the NJOY 99.259 system are reported.