Models of Ceramic Tritium Breeding Zones in the Blanket of a Thermonuclear Reactor

The development, performed in the 1980s–1990s, of models of tritium breeding zones for blankets of thermonuclear reactors, based on the use of ceramic lithium-containing materials, is described. 5 figures, 1 table.     

Corrosion resistance of fuel element steel cladding in a lead coolant

The corrosion resistance of experimental and industrial steels considered as possible materials for fuel element cladding in the BREST reactor is investigated. Static corrosion tests in lead at 750°C lasting up to 1000 h, as well as metallographic and x-ray analysis, have been performed. Data are provided on the composition and relative width of a diffusion barrier in the form of complex oxides arising in the corrosion interaction zone and impeding the diffusion of oxygen into steel, as well as the values of the averaged oxygen diffusion coefficient. The corrosion resistance in lead of the newly developed steel Kh18S2VBFMAYu and steel 16Kh12MVSFBR (ÉP 823) proposed as cladding for fuel elements in the BREST reactor is analyzed.     

A Theoretical Study of the Kinetics of Hydride Cracking in Zirconium Alloys

A diffusion model is suggested for computing the rate of delayed hydride cracking (DHC) in nonirradiated zirconium-base alloys. The rates of DHC in claddings of fuel elements of VVÉR and RBMK reactors and in pressure pipes of CANDU reactors are predicted.     

Methods for protecting fuel-element cladding from corrosion in lead coolant and the heat-transfer sublayer

The results of investigations of the corrosion of commercial and experimental steels in lead and the possibilities of corrosion protection are presented. The effect of lead coolant and the lead heat-transfer sublayer on fuel-element cladding are examined. Methods based on thermodynamic calculations and experimental data are proposed for protecting fuel element cladding in a lead-cooled reactor from the corrosive effect of the coolant by creating a new corrosion resistant chromium steel and from the corrosive effect of the heat-transfer sublayer by alloying with the components of steel. The results of this work have been implemented in the experimental fuel elements for the BREST-OD-300 reactor which were irradiated in a BOR-60 reactor. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 88–94, February, 2008.     

Analysis of the operational reliability of VVÉR-1000 fuel elements and bundles in a three-year fuel cycle

Institute of Nuclear Reactors, Kurchatov Institute, Russian Science Center. A. A. Bochvar All-Union Scientific Research Institute of Inorganic Materials. All-Union Scientific Research Institute of Nuclear Power Plants. Novo-Voronezh Nuclear Power Plant. Production Association, Novosibirsk Chemical Concentrate Factory. Translated from Atomnaya Énergiya, Vol. 74, No. 6, pp. 482-486, June, 1993.