A. M. Tsibulya

An Overview of the Experiments Performed at the BFS Facilities and Evaluated for the International Reactor Physics Experiment Evaluation Project

Abstract In total, eight evaluations with 27 critical configurations spanning 11 series of assemblies conducted at the BFS-1 and BFS-2 facilities [BFS: Bol’shoy Fizicheskiy Stand (in transcription from Russian: Big Physical Facility)] are documented in the 2013 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). The evaluations concern four scientific issues: physics of sodium-cooled, medium-sized fast reactors; physics of sodium-cooled, small-sized experimental reactors; physics of lead-cooled fast reactors; and integral benchmark data for the mixed oxide (MOX) fuel manufacturing process, particularly low-moderated MOX fissile media. Also, two evaluations with 11 critical configurations spanning two series of assemblies conducted at the BFS-1 facility where many physical parameters in addition to criticality were measured and are documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). These experiments were designed to obtain data that are applicable to a wide range of criticality safety operations involving fissile contaminated waste streams. This paper provides an overview of the BFS experiments presented in the IRPhEP and ICSBEP Handbooks.

89Sr PRODUCTION IN FAST REACTORS

ConclusionsIn summary, we have proposed a new method for producing89Sr for medical purposes from natural yttrium according to the reaction (n, p) in fast-neutron reactors. Investigations confirm the computational parameters of the production: from 2 to 15 mCi89Sr per gram of the starting yttrium. We have shown that90Sr can be extracted and the final product with the required radionuclide purity can be obtained. Commercial production of89Sr in BR-10 and BOR-60 has now started.

Determination of the237Np(n, 2n) reaction cross section in the core of the BN-350 reactor

Analysis of irradiated fuel (regular fuel elements and special ampule samples) makes it possible to obtain data on the neutron-physical characteristics that do not lend themselves to measurement in experiments with a low neutron flux. The authors studied samples of spent fuel from regular fuel elements and samples of /sup 236/U and /sup 237/Np irradiated in special ampules. We give the results of measurements of the ratio (sigma/sub n,2n/ + sigma/sub ..gamma..,n/)sigma/sub c/ for /sup 237/Np in different zones of the BN-350 reactor and the results of calculations performed using the neptunium group constants calculated from the data of the files ENDFB IV and ENDFB V

Determination of the cross sections for some neutron reactions with structural materials in the BN-350 reactor core

The authors experimentally and theoretically determine the differential cross sections for some neutron reactions with the target nuclei of several BN-350 reactor material constituents. The nuclei include iron 54, iron 56, nickel 58, molybdenum 92, cobalt 59, manganese 55, and uranium 235. The cross sections were determined relative to the uranium 235 fission cross section by fast neutrons. Irradiation procedures are discussed and results are tabulated. The spatial distribution of the neutron and gamma spectra was calculated using the six-group approximation and the BNAB-78 system of constants.