No abstract available

No abstract available

[en] The reliability of critical safety evaluation of MOX fuel management is determined by ²³⁹Pu and ²³⁸U neutron data accuracy. A qualitative analysis of up-to-date knowledge on these data, important for nuclear safety assurance, is given in the paper. The analysis of existing experimental information shows a need for additional integral benchmark experiments for low-moderated MOX media. Experiments performed and planned at the BFS facility might provide valuable information on the quality of nuclear data. (author)

No abstract available

[en] This paper presents an overview of activities performed to prepare and test the group constants ABBN-90. The ABBN-90 set is designed for application calculations of fast, intermediate and thermal nuclear reactors. The calculations of subgroup parameters are discussed. The processing code system GRUCON is mentioned in comparison to the NJOY code system. Proposals are made for future activities. (author). Figs, tabs

[en] At present the commercial fast sodium cooled reactor (BNK) is under development in Russia. Initially electric power output of this reactor was chosen 1800 MW(e). However, further power output was decreased down to 1200 MW(e) to provide transportation of the main equipment by rail. The main concept of the core for this reactor was taken from BN-1800 reactor: low core specific power, internal self-protection (close to zero sodium void reactivity effect (SVRE) value, decreased reactivity margin for fuel burn-up). This paper presents the results of theoretical and calculation studies on choosing and optimizing physics parameters of BNK-1800 type reactor core and BNK-1200 type reactor core more detail. There is a set of possibilities for improving the core for BNK-1200 type reactor, staying within limits of new design. These possibilities are to improve flattening of the core power field, to provide close to zero value of reactivity margin for fuel burn-up and other. Unique enrichment of fuel and flattening of the power field by steel absorbers was optimal solution for BNK-1800 core with diameter of above 6 m, but for BNK-1200 core of smaller dimensions flattening of power field by two enrichments allows an essential decrease (down to 10%) of maximum specific power and maximum fuel burn-up (at the same average fuel burn-up). In 2008 in Russia Nuclear Safety Rules (PBya RU AS) had been changed. The requirement of negative reactivity coefficient on coolant density was removed. Concerning SVRE, new Rules state the following: the interval of allowable positive SVRE values should be defined in the design of Reactor Installation. It allows to extend the area of optimal values of the core parameters and, in particular, to increase the core height up to 100 cm. It is possible to realize it at the expense of decreasing sodium plenum dimension. Increase of the core height (with corresponding decrease of its radial dimension) leads to essential increase in efficiency of CSS rods that, in principle, allows to avoid the use of expensive high-enriched boron in CSS rods. Moreover, the increase of the core height simplifies the problem of providing zero reactivity margin for fuel burn-up. For this purpose, it is necessary to increase fuel volume fraction (up to ∼ 0.53) and to increase the number of SAs within the limits of design with the aim to avoid the hydraulics problems. In principle, introduction of the upper blanket above the core allows to reach breeding ratio (BR) value of more than 1.4 as in the case with nitride fuel with old fuel volume fraction (∼0,4). Critical loading of the core together with BR is the principle issue for nuclear power. It can restrict the pace of fast reactor introduction. Decrease of the critical loading is possible only at the expense of increasing core specific power. Comparison of consumption of plutonium amounts in the cores with low specific power and high specific power with external fuel cycle duration of 3 years showed that the benefit of high specific power core concept by plutonium consumption has been observed only during 5-7 years, and this benefit has been lost on the large time interval. (authors)

[en] The ABBN-93 299-group cross-section library is used as a source of neutron data for criticality calculations with the MMK-KENO Monte Carlo criticality code. Validation of criticality calculations is performed using the data presented in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. This paper contains a description of a validation method based on statistical analysis of discrepancies between calculated and benchmark-model k_eff's and the results of this validation for different types of experiments. Another validation method using a well-known procedure of group cross-section adjustment based on the maximum likelihood method (generalized least-squares method) and results of the validation for water-moderated highly enriched uranium homogeneous critical systems using selected experiments of the HEU-SOL-THERM type are also presented

[en] The necessity of accounting for the correlation of experimental uncertainties when determining calculational uncertainty in criticality predictions is demonstrated in this paper. An example of constructing the correlation matrix for the experimental uncertainties of the experiments with highly enriched uranium solutions performed and evaluated at the Institute of Physics and Power Engineering, Russian Federation, is presented. A conclusion is drawn with regard to the significant influence of the correlations on the average calculation-experiment deviation and the deviation uncertainty

[en] The library FOND-2.2 of evaluated nuclear data files, which was created at the ABBN laboratory of IPPE, is described. FOND-2 library is the basic nuclear data source used for the preparation of group data sets with different energy structures. ABBN-93.1 group data set was retrieved from the FOND-2 data library and nowadays it is widely used in different applications, in neutronics calculations of different nuclear energetic installations with different kinds of neutron spectra, in radiation shielding calculation, and so on. (author)

[en] Efforts in the field of experimental and calculation studies of BREST-OD-300 reactor physics in the whole are focused on: Creation of nuclear constant data and calculation software verified and certified by GAN RF; Justification of reliability and precision extent of calculation prediction of main neutron-physical characteristics of BREST-OD-300 reactor core; Confirmation of nuclear and radiation safety of BREST-OD-300 reactor facility and plant fuel cycle; Correction and refinement of design parameters, control and regulation systems, composition of starting and subsequent critical loadings, planning of reloading schedules. The main directions of activity - development of nuclear constant data and calculation software, verification and certification, experiments at BFS stand, calculation analysis of experimental results, elaboration of international benchmark models of BREST-OD-300 reactor, estimation and analysis of prediction precision for the main neutron-physical parameters of BREST-OD-300 reactor. The report reviews the main activities on the above mentioned directions, gives a number of results obtained, and concludes on the compliance extent of required and achieved calculation precision of neutron-physical characteristics of BREST-OD-300 reactor. (author)