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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.

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.

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.