XIX-th ARS SEPARATORIA – Złoty Potok, Poland 2004
SOME ISOTOPE SEPARATION PROBLEMS IN SPENT
FUEL PROCESSING
V.D. BORISEVICH, V.I.PETROV, G.A. SULABERIDZE and Q. XIE1)
Moscow Engineering Physics Institute
(State University)
Moscow 115409, 31 Kashirskoe Shosse
1)
Permanent postal address:
Research Institute of Physical and Chemical Engineering of Nuclear Industry,
Tianjin 300180, P.R. China.
INTRODUCTION
Impossibility to burn down nuclear fuel for a single stay in a reactor of
an atomic power station (APS) dictates need of its multiple recurrence to a
process and hereunder close-up a fuel circulation in a nuclear fuel cycle
(NFC). Only under these circumstances it will provide the decision of a
strategic purpose of a NFC usage: provision of the nuclear energetics with
fuel for a long-term perspective and simultaneous solution of the treatment
problem for the nuclear wastes appeared.
The fall in rate and scale of development of the nuclear energetics
after the Chernobyl disaster has reduced urgency of transition to the closed
NFC. However, the last trends in the world energetics make clear its
temporary character. In the future, the nuclear energetics can not be efficient
without turning on the closed fuel cycle both in consequence of inevitable
raise of a price of natural uranium and need of conversion of unloaded
(spent) fuel because of a high price of its constant storage. The estimations
show that repeated usage of spent fuel of nuclear reactors leads to a
noticeable saving of natural raw material. Besides, reprocessed uranium
contains more uranium-235 than natural one (in the limits of 0.8-2.5 weight
% depending on a type of a nuclear reactor). So re-enrichment of
reprocessed uranium, say, to the grade necessary for Light Water Reactor
(LWR) fuel will lead to reduction of the charges on the separative work.
PROBLEM STATEMENT
The repeated use of the recovered uranium faces two main problems
that the separation industry never met during enrichment of natural uranium.
First of than is connected to the fact that burning-out of the nuclear fuel of
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energy nuclear reactors is producing a uranium-232 nucleid, in which a
chain of a radioactive decay contains various long-lived radioactive
elements, in particular tallium-208, that are the sources of hard gammaradiation creating high radioactive background [2]. In time, a gammaactivity of spent fuel increases significantly, reaching its balance value after
approximately 10 years of storage [3-4].
It is also necessary to keep in view that enrichment of reprocessed
uranium in a separation cascade causes in its product flow a concentration of
the uranium-232 isotope, which will be enriched together with the key
isotope uranium-235. As a result of such a process of re-enrichment, the
difficulties with provision of radiation safety either for personnel or for
separation equipment (defined by accumulation of radioactive products of
uranium-232 decay) will appear. This circumstance poses a problem of
content restriction of uranium-232 in recovered nuclear fuel. It was revealed
that the ratio of uranium-232 concentration to that of uranium-235 in
recovered fuel must not exceed ∼1×10-7 [4].
The second problem is connected to the situation that in the LWR
active zone a big amount of uranium-235 does not participate in the fission
reaction, and being captured by a nuclei of uranium-235 transforms into nonfissionable uranium-236. This nucleid is a parasitic neutron absorber that
effects negatively on a reactor reactivity and a depth of its burning-up. The
fresh fuel containing uranium-236 must be enriched more with uranium-235
than free from uranium-236 in order to maintain an equal energy output per
unit of uranium. The compensation factor is defined as a ratio of the
additional uranium-235 enrichment required to compensate for the reactivity
penalty due to uranium-236 to the concentration of uranium-235. Usually it
evaluates within the limits of 0.2÷0.6 [3, 5-6].
Besides, in the case of multiple returns of recovered uranium to a fuel
cycle a concentration of uranium-236 will constantly grow with a
simultaneous increase of the separation work required for production of
given condition of nuclear fuel i.e. a fuel with necessary enrichment of the
fissionable uranium-235 isotope.
Basing on the above discussion, the main separation problems in reenrichment of irradiated fuels on uranium-235 can be formulated as follows:
1. determination of special cascade schemes for re-enrichment of
uranium-235 with simultaneous decontamination of a fuel from the
uranium-232 isotope or diluting isotope composition to
concentration of this isotope up to acceptable value;
2. searching optimum ways to re-enrich recycling uranium with
uranium-235 with taking into account necessary compensation of
uranium-236 influence;
3. choice of cascade schemes for recovering uranium-235 to a
necessary level with simultaneous decontamination of reprocessed
fuel from the isotopes of uranium-232 and uranium -236.
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In this study the enumerated isotope separation problems are
discussed with account of the last two decades achievements both in Russia
and abroad in the theory of isotope separation in cascades. To describe the
mass transfer process of multicomponent isotope mixtures are used so-called
R-cascades (or in other word quasi-ideal cascades for separation of multiisotope mixtures) with large enrichments at their steps. The theory of such
cascades has been developed in [7-12].
RESULTS OF RESEARCH
Table 1 shows a typical isotopic composition of uranium in the
irradiated fuels of LWR reactors after 10 years of exposure [1].
Table 1. Typical isotopic composition of uranium in irradiated fuels of LWR
reactors after 10 years of exposure
Uranium
isotope
232
234
235
236
238
Concentra-tion,
% wt.
(0.9÷2.5)×10-7
∼2.3×10-3
0.8÷1.2
0.45÷0.65
the
rest
In the research carried out it has been assumed that re-enrichment of
recycling uranium has to provide a concentration of uranium-235 not less
than 3.5 weight % and ratio of the uranium-232 concentration to the
uranium-235 concentrations must not exceed 1.1×10-7. Besides, the
additional uranium-235 enrichment to compensate presence of uranium-236
is defined by the compensation factor varying within the range of 0.2 up to
0.6. The uranium-235 concentration in the waste flows of the separation
cascades under investigation was assigned as equal to 0.2 weight % [1].
The theoretical analysis has shown that re-enrichment of recycling
uranium with simultaneous reduction of uranium-232 and uranium-236
isotope concentrations can be realized with the help of one of the schemes
presented in Fig.1 The mixture of recycling uranium isotopes can be diluted
by the natural raw material in three different ways: in the flow entering
cascade (so-called the feed flow, Fig.1а), in one of the flows leaving cascade
(the product flow, Fig.1b). Besides, it is possible to feed up a cascade with
the flow of recycling uranium as an additional feed flow in some
intermediate step of a separation installation (Fig.1c).
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XIX-th ARS SEPARATORIA – Złoty Potok, Poland 2004
Fnat
Fnat
Freg
P
W
Freg
W
P
a
b
P
Fnat
Freg
Fig.1. The possible R-cascade schemes
to re-enrich uranium-235.
W
P
c
In this study it is demonstrated that the cascade scheme presented in
Fig. 1c is the most advantageous form the standpoint of saving the separative
work. So, the paper examines the aspects of re-enrichment of uranium-235
by the cascade scheme with the additional feed flow.
SUMMARY AND CONCLUSIONS
The calculations of uranium isotope compositions in a product flow of
a chosen separation cascade, the specific separative work Аspec., the factor of
natural uranium consumption defined as f 0 = Fnat /P , the ratio flows of
natural to reprocessing uranium Fnat /Freg versus the factor of reprocessing
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uranium consumption freg = Freg /P , were carried out. It was demonstrated
that diluting uranium-232 demands keeping certain correlation between
values f0 and freg. Moreover, the f0/freg ratio depends on the isotopic
composition of reprocessing uranium, desirable enrichment of the uranium235 isotope as well as maximum possible concentration of uranium -232.
As it follows from the results obtained, the value of the specific
separative power Аspec. decreases with an increase of the concentration of
uranium-235 in the reprocessing uranium flow Freg and under these
conditions saving of the separative work to the work required for enrichment
of natural uranium to the same concentration of uranium-235 can achieve
from 10 to 15% with saving of natural raw material up to 20%.
The additional expenses of the separative work, because of necessity
to compensate the presence of uranium-236 in the product flow for 1%
concentration of uranium-235 in reprocessing flow, will be in the limits of
(9-13)% depending on the chosen compensation factor.
We are obligated to say that the results of this paper have the
preliminary nature. The final conclusions on the optimum strategy of
conversion of spent fuels must be executed with more economic indices,
characterizing cost of transportation of downloaded fuels, its chemical
conversion, costs of the separation work unit and necessary radiation
protection, and as well as some other components forming the final price of
the recovered product.
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