Correcting the isotopic composition of regenerated uranium with respect to 232U by a centrifuge method with introduction of a carrier gas

Abstract

Theoretical and experimental studies dealing with correcting the isotopic composition of regenerated uranium with respect to 232 U by a centrifuge method with introduction of a carrier gas are reported. In order to increase the efficiency of separating 232 U from the spent uranium and reduce the loss of 235 U, the use of a carrier gas is proposed -the gaseous compound 12 C 8 H 3 F 13 , which is inert to uranium hexafluoride, and whose molecular weight, M c = 346 amu, matches that of 232 UF 6 . Freon, C 8 H 3 F 13 , is shown not to decompose during operation in the rotor of a centrifuge or to interact with the centrifuge material. The measured absorption parameters of freon on sodium fluoride NaF confirm the feasibility of efficient separation of a mixture of uranium hexafluoride and freon with return of the freon to the separation process. It is shown that introducing a carrier gas into the centrifuge technology can yield some new results: lowering the radioactivity of the commercial product, normalizing the overall radiation situation during production, increasing the recovered 235 U in the commercial product, and reducing the volume of radioactive waste. The recovery of 235 U in the commercial product can be increased to 99% or more. Then the 232 U content in the commercial product is ~2·10 -8 % or a factor of 10 less than the maximum allowable content of 2·10 -7 %.Correcting the isotopic composition of regenerated uranium with respect to 232 U is an especially important task in connection with the anticipated expansion in the amount of reprocessing of spent nuclear fuel for the closed fuel cycle for large scale nuclear power generation [1,2].The purpose of this paper is to study the problems at a separations plant owing to the isotopic composition of uranium in spent nuclear fuel from nuclear power plants using light water reactors.The use of regenerated uranium can lead to significant savings of natural uranium and economy in separation processing, since the amount of 235 U in spent fuel from VVER or PWR reactors is higher than in natural uranium ( Table 1). The regenerated product contains 232 U (~10 -7 %), whose decay chain includes 208 Tl which is a high energy γ emitter, 234 U (~0.05%), which is an intense α emitter, 235 U (~1%), and 236 U, which is a strong neutron absorber.The presence in the spent fuel of 232 U, with its highly active γ-emitting decay products, makes it more difficult to use regenerated uranium in a closed nuclear fuel cycle. The high γ activity of the daughters of 232 U makes regenerated uranium unsuitable for use in modern nuclear fuel preparation technology.

Purification and Enrichment of Regenerated Uranium in a Double Cascade

“…The purifi ed uranium hexafl uoride waste of the second cascade is diluted to the require 235 U concentration. In the method where an additional carrying gas is used as the feed fl ow of this cascade, the 232 U content decreases to 3·10 −8 -4·10 −9 % without the use of a special diluent [8]. A double cascade has the drawback that high-enrichment…”

“…uranium is obtained with 235 U content above 20% at separate stages of separation [6][7][8]. An alternative is a double cascade with a limitation on the 235 U concentration and an altered arrangement, where the purifi cation by 232,234 U removal is accomplished in the fi rst (ordinary) cascade (see Fig.…”