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[en] The description is given of two program modules - BRAM and MES H2 which are designed for preparing the files of data being used for starting up of reactor cell calculation programs. The module BRAM contains a basic library of reactor materials (only nuclear-physical compositions) and also allows to form a temporary library of materials. The module MES H2 realizes preparing the data files for reactor cell calculation by means of WIMSD4 code. Under conditions of dialogue with a user on the basis of temporary library of materials it is given the geometry and cell composition with graphical representation and editing, the choice of calculation method, the specification of various options

[en] The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for 1) Critical approach and Exponential experiment, 2) Measurement of neutron flux distribution, 3) Measurement of power distribution, 4) Measurement of fuel rod worth distribution, and 5) Measurement of safety plate worth by the rod drop method. (author)

[en] The nuclear design, calculating model and codes for the first pulsed type research reactor in China are outlined. The comparison between the calculation results and the zero power experiments is given also

[en] A nodal integral method is derived for the monoenergetic, steady-state, fixed source neutron diffusion equation in hexagonal geometry based on a coordinate transformation that maps a parallelogram into a rectangle. The new hexagonal nodal diffusion method is implemented in the computer code HND where the discrete-variable equations are solved via an iterative scheme. Because the new method's equations are derived for a rhombus, they can be solved on a sequence of embedded meshes to study the method's error order. Indeed a preliminary numerical error analysis reveals a second-order error in the mesh size, and comparison with finite difference results obtained with the finite difference based BOLD-VENTURE code indicate the superior accuracy of our new nodal method

[en] A constructive method for estimating asymptotic stability domains of nonlinear reactor models is developed. The method is based on expansion of a positive definite function instead of a Lyapunov function and thus it is easy to come up with a starting region. The method starts with the region given by a specific positive definite function and provides a sequence region, the entire stability region is estimated effectively after sufficient iterations. It is particularly useful for reactor systems that are stiff. The method is applied to several nonlinear reactor models in the literature and the results are compared

[en] The GETERA code is intended for calculation of neutron space-energy distribution in nuclear reactor cells and polycells and various neutron flux functionals by the collision probability method in multigroup approximation. The code uses the 26-group BNAB nuclear data library for slowing down diapason. In the 0 to 2.15 eV thermalization range the code uses a special 100-group neutron cross-section library based on the ENDF/B-IV and JENDL-2 evaluated nuclear data files

[en] The present report deals with a material which is used as a neutron absorber in High Temperature Graphite Reactors. There were no multigroup data for Hafnium in the thermal and epithermal libraries of VSOP code. On the other hand, natural Hafnium is one of the materials for which ENDF/B-VI data are now available. Neutron absorption by Hafnium occursat the thermal energies, and in many resonances, most of which are above the thermal energy region. The preparation of multigroup libraries for natural Hafnium in the VSOP format from the ENDF/B-VI basic data is discussed. (author)

[en] Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the standard for use in k_eff calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, σ_p, for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of σ_p to characterize resonance self shielding. Three prescriptions for calculating σ_p are given. Finally, results of several calculations of k_eff on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems. (Author)

[en] Short communication