Vol 126, No 3 (2019)

The issues of Np, Am, Cm recycling in a BZhSR molten-salt reactor with a fast neutron spectrum based on LiF–NaF–KF eutectic are examined. The results of a comparative analysis of the efficacy of transmutation for different BZhSR fuel compositions and core volume are reported. It is shown that BZhSR with an initial load of enriched uranium and Np, Am, Cm fl uorides makes it possible to transmute ~1 ton/yr Np, Am, Cm on average over 50 years of operation at 1650 MW(t). The advantage of the chosen arrangement is the capability of Np, Am, Cm replenishment without adding additional enriched uranium even after the first run with duration 300 EFPD.

Thermohydraulic validation is a priority for the ZhSR-S [MOSART] molten-salt nuclear transmuter for transuranium elements from the spent nuclear fuel of solid-fuel thermal reactors. In the present article, the results of coupled neutronics and thermohydraulics calculations, based on which the configuration of the cavity core and fuel loop were chosen, are generalized. The thermohydraulic validation of such a reactor was preceded by the development of an experimental database on the physical properties of the fuel salt with the molar composition 73LiF–27BeF₂ + AnF₃ and 58NaF–15LiF–27BeF₂ + AnF₃. The optimized 2400 MW(t) ZhSr-S core satisfies the two most important requirements of thermohydraulics: no recirculation or stagnant zones and quite low maximum temperature of the solid reflectors so that they can be used for a long time.

The aim of this work is to develop methods for calculating the neutronic parameters of HTGR and to perform uncertainty analysis with the aid of design programs and neutron covariance data. The method and software modules for preparing multigroup libraries of covariance data for individual isotopes based on a 44-group library of covariance data of the SCALE-6 computational complex were improved in the course of this work. Correction factors are introduced to improve the accuracy of the method. The WIMSD-5 code with a 69-group cross-section libraries, created on the basis of ENDF/BVII.0 and ENDF/BVII.1 evaluated nuclear data files, is used for the neutronic calculations. Calculations of the uncertainties of one-group fission and absorption cross sections for the main isotopes of the elements of the MHTGR-350 core as well as the neutron multiplication coefficients for the fuel-compact cells and fuel-block cells of the MHTGR-350 reactor were performed to verify the developed method.

Reactor tests in BN-600 and post-reactor studies of experimental FA of fuel rods with mixed nitride fuel and different cladding are being conducted in Project Breakthrough. In the present article, the irradiation parameters of fuel rods with mixed nitride fuel and cladding made from ChS68-ID c.d. austenitic steel are presented: maximum linear power density 38.3 kW/m, maximum fuel burnup 7.5% h.a., and maximum damage dose 73.6 dpa. No depressurization of the fuel elements was recorded. The results of post-reactor examination (PIE) of the irradiated fuel rods are presented – swelling, gas release from fuel, and mechanical and corrosion characteristics of the cladding.

According to the Szilard–Chalmers effect, ⁹⁹Mo recoil atoms can be obtained in nuclear reactions and recorded in a collector. Knowledge of the dependence of the yield of ⁹⁹Mo atoms on the thickness of the molybdenum layer is necessary for their efficient collection. The yield of ⁹⁹Mo recoil atoms from molybdenum nanolayers in the nuclear reaction ¹⁰⁰Mo(p, x)⁹⁹Mo was measured as a function of the thickness of the nanolayer. Nanolayers of metallic molybdenum were fabricated by magnetron sputtering on sapphire plates. The measurements were performed after the nanolayers were irradiated by 28 MeV protons in the U-150 cyclotron. The yield of ⁹⁹Mo recoil atoms for 38–205 nm thick nanolayers was 65–8%. It was found that the maximum ⁹⁹Mo yield obtains with molybdenum layer thickness 80 ± 5 nm. It was found that the free path of ⁹⁹Mo recoil atoms in native metallic molybdenum is equal to 34 ± 9 nm.