The production of vibrationally c o m p a c t e d fuel e l e m e n t s f r o m granulated carbide and nitride fuel is also of i n t e r e s t . The technology of obtaining g r a n u l a t e d c a r b i d e - - n i t r i d e fuel with the r e q u i r e d density and h o m o genity of distribution of u r a n i u m and plutonium is still in the development stage. The p r e l i m i n a r y data enable us to hope that the s o l - g e l p r o c e s s e s and the c r y o c h e m i c a l method m a y e n s u r e the production of u r a n i u m - plutonium c a r b i d e nitride fuel with c h a r a c t e r i s t i c s n e c e s s a r y for the production of fuel e l e m e n t s for fast power breeder reactors. The above i n f o r m a t i o n shows the extensive p o s s i b i l i t i e s of using different technological f a c t o r s f o r controlling the quality of obtained oxide and carbide nitride c o r e s of fuel e l e m e n t s . As the e x p e r i m e n t data a c c u m u l a t e s , the r e q u i r e m e n t s i m p o s e d on the n u c l e a r fuel will be refined, which could c e r t a i n l y lead to an i n c r e a s e in its efficiency and reliability of fuel e l e m e n t s . The i m p r o v e m e n t and i n c r e a s e d s t r i n g e n c y of the r e q u i r e m e n t s on n u c l e a r fuel is a n a t u r a l p r o c e s s . However, these r e q u i r e m e n t s m u s t be sufficiently justified so that the cost of the fuel e l e m e n t s as a whole does not i n c r e a s e unjustifiably and the economy of operation of a t o m i c power plants is e n s u r e d . The total p r o j e c t e d p o w e r of a t o m i c p o w e r plants in the e n t i r e world in the n e a r future exceeds 100 MW (electrical) [4]. T h e s e a r e mainly installations with t h e r m a l r e a c t o r s with fuel e l e m e n t s made of u r a n i u m oxide in z i r c o n i u m c a s i n g s . E x p e r i e n c e in the operation with t h e s e fuels will facilitate subsequent t r a n s i t i o n to the mixed u r a n i u m - - p l u t o n i u m oxide fuel in t h e r m a l and fast r e a c t o r s ; a continuation of the r e a c t o r i n v e s tigations of c a r b i d e - - n i t r i d e fuel will p e r m i t a m o r e c o m p l e t e understanding of its potentials and will a c c e l e r a t e w i d e s p r e a d u s e of this p r o m i s i n g fuel in fast r e a c t o r s . LITERATURE 1. 2.

3. 4.

CITED

F. G. Reshetnikov et a l . , in: P r o c . IAEA Syrup. "Fuel and Fuel E l e m e n t s f o r F a s t R e a c t o r s , " Vienna (1974). G. K a r s t e n , T r a n s . Am. Nucl. S o c . , 19, No. ! , 84 (1974). R. H i l b e r t e t a l , T r a n s . A m . Nucl. Soc., 19, No. 1, 135 (1974). Atomic Engineering Abroad, No. 12, 12 (1973); No. 2, 16 (1974); No. 9, 11 (1975).

PLUTONIUM

FUEL

FOR

REACTORS

POWER

AND

FUEL

ELEMENTS

I. S. Golovnin, A. S. Zaimovskii, T. S. Men'shikova, N. P . A g a p o v a , Yu. K. Bibilashvili, V. A . T s y k a n o v , E. F. Davydov, V. M. Gryazev, V. I. Kuz'min, V. r . S y u z e v , V. M. Sedov, and N. P. Dergachev

A c o m p a r i s o n of the growing demand of e n e r g y with the g u a r a n t e e d availability of u r a n i u m o r e s , even considering opening of new natural r e s o u r c e s and e x t r a c t i o n of uranium f r o m s e a w a t e r , shows that the p r o b lem of n u c l e a r fuel can be solved by introducing f a s t r e a c t o r s along with the r e l a t i v e l y cheap t h e r m a l r e a c t o r s , which as a r e s u l t of extensive breeding of s e c o n d a r y n u c l e a r fuel would p e r m i t using p r a c t i c a l l y all obtainable u r a n i u m [1]. Thus, in the complex fuel cycle of future nuclear e n e r g e t i c s , one m u s t provide for a wides p r e a d u s e of u r a n i u m - - p l u t o n i u m fuel both in fast and t h e r m a l nuclear r e a c t o r s . One of the main r e q u i r e m e n t s on atomic p o w e r plants with fast r e a c t o r s and on r a d i o c h e m i c a l production is to e n s u r e a r a t e of b r e e d ing of plutonium that would p e r m i t doubling the r a t e d p o w e r output in not l e s s than 10 y e a r s . In n u c l e a r e n e r g e t i c s the use of plutonium fuel is c u r r e n t l y being investigated along with the economic a s p e c t s of the p r o b l e m in w a t e r and w a t e r - - g r a p h i t e r e a c t o r s . In fast r e a c t o r s u r a n i u m - - p l u t o n i u m fuel is a l r e a d y being u s e d , since the use of 235U in t h e m is not p r o m i s i n g because of the low r a t e of breeding. The knowledge of the p h y s i c o m e c h a n i c a l and the technological c h a r a c t e r i s t i c s of u r a n i u m and plutonium oxides, t h e i r c h e m i c a l and radiation s t a b i l i t y , and a l s o the r e l a t i v e s i m p l i c i t y of technological methods of producing u r a n i u m - - p l u t o n i u m fuel e l e m e n t s led to i m m e d i a t e development of just this type of fuel. T r a n s l a t e d f r o m Atomnaya ]~nergiya, Vol. 43, No. 5, pp. 412-417, N o v e m b e r , 1977.

0038-531X/77/4305-1063 $07.50 9 1978 Plenum Publishing C o r p o r a t i o n

1063

50

m

ZO t~

10

0

q

lZSO

tSO0

I

1750

~

Fig. 1 Fig. 2 Fig. i. Dependence of completeness of dissolution of brickets of mechanically mixed powders of UO 2 and PuO 2 (O) and of chemically coprecipitated powders ~qJ, Pu)O 2 (e) in nitric acid, on their baking temperature. Fig. 2. Microstructure of core of (U, Pu)O 2 obtained by baking of mechauieally at a temperature of 1750~ for 3 h (x 340).

mixed

powders

Investigations have shown that in the future the widespread use of mixed oxide fuel in thermal watercooled reactors can be foreseen. But in fast reactors the use of such fuel is apparently limited to the first and second generations of power installations. Oxide fuel with relatively high breeding coefficients (~ !.4) cannot ensure (considering the entire fuel cycle) the projected growth of power due to the low rate of accumulation of plutonium or large technical difficulties in attaining required rate of accumulation. However, the initial industrial use of this type of fuel in fast reactors has enormous sig~2ficance for establishing largescale production of mixed oxide fuel for thermal reactors. Although the above question is subject to further study, the problem of developing more efficient forms of fuels is undoubtedly a pressing one; this primarily refers to the mixed uranium--plutonium carbide and nitride fuel whose physical characteristics determining the rate of breeding of secondary fuel are better than those of oxide fuel. However, the technological characteristics of these materials are worse and the operational characteristics of the fuel elements made from them have not been adequately investigated. Characteristics of Fuel Elements of Power Reactors with (U, Pu)O 2 Fuel. In the development of fuel elements made of uranium--plutonium oxide fuel, a number of characteristics related to the use of this composition must be taken into consideration; among these are: i. The decrease of melting temperature of the mixed fuel compared to uranium. This has certain significance only for fast reactors. For the fuel of thermal reactors containing 3-6% PuO2, the decrease of melting temperature of the core can be neglected. Here the higher self-screening of the neutron field, caused by the presence of plutonium, not only considerably lowers the gas release but also permits either increasing the melting margin or increasingthe thermal loading on the fuel element by 10-15%. 2. The increase of the oxygen potential in the fuel element with the increase of depletion. A consequence of this is the increased physicochemical interaction of the fission products with the fuel element shells. The solution of this problem is connected with the determination of the optimum initial oxygen coefficient of the fuel core and with the development of a technological process which would permit preparing fuel element cores with given oxygen coefficient. 3. The uniformity of distribution of plutonium in the fuel core. This condition is one of the important safety criteria for fast reactors. Physical investigations have shown that in the mixed fuel the admissible size

1064

6

O 2 0

v

:om 100 200 J90 480 580 Length of the active part of the fuel element, mm Fig. 3 Fig. 4 Fig. 3. Vacancy p o r e s and s e g r e g a t i o n of Nb(C, N) in the fuel-element shell made of 0Khl6N15M3B steel for E > 0.1 MeV and T i r r ~ 4 5 0 (a) and T i r r ~ 5 0 0 (b) ~ (x 200,000). Bot

Fig. 4. Change of density of shell along the length of the fuel element: O) experiment; o) c o m putation. of P u O 2 p a r t i c l e s must not exceed 50 p. T h e r e f o r e , on the one hand, the technology of p r e p a r a t i o n of fuel c o r e s must e n s u r e the required homogenity and, on the other hand, it must be considered that during the o p e r ation s e g r e g a t i o n of plutonium o c c u r s in the fuel c o r e due to high t e m p e r a t u r e gradients. 4. The transition to plutonium fuel in t h e r m a l r e a c t o r s demands optimization of individual p a r a m e t e r s , _primarily the depth of depletion and t h e r m a l s t r e s s e s of the corresponding construction of the fuel elements. In using plutonium these p a r a m e t e r s must i n c r e a s e , since the cost of p r e p a r a t i o n of the fuel elements i n c r e a ses sharply. 5. In using plutonium fuel in power r e a c t o r s , it is required to r e p r o c e s s a v e r y l a r g e amount of highly toxic n u c l e a r fuel. This is especially true for t h e r m a l r e a c t o r s (amount of r e p r o c e s s e d fuel in these is ~ 10 t i m e s l a r g e r than in fast power r e a c t o r s of the s a m e capacity). Methods of Obtaining Cores of (U, Pu)O a. The c o r e s of fueI elements for fast r e a c t o r s can have the f o r m of t a b l e t s , s l e e v e s , or vibrationally compacted powder with the content of PuO 2 in the mixture ranging f r o m 10 to 30%. At present the f i r s t two types of oxide c o r e s obtained by the methods of powder m e t a l l u r g y a r e widely used. As the initial powder, powders p r e p a r e d by mechanical mixing o f UO2 and PuO 2 powders (the formation of homogeneous solid solution of (U, P u ) O 2 o c c u r s during baking at a high t e m p e r a t u r e ) and by chemical coprecipitation of nitric solution of uranium and plutonium, when the original powders a r e a solid solution of (U, Pu)02, a r e used. The g r e a t e s t difficulties a r e encountered in using mechanically mixed powders. In o r d e r to meet the specified r e q u i r e m e n t s on the density and uniformity of distribution of plutonium, a p r e l i m i n a r y t r e a t m e n t of the powders and a sufficiently high baking t e m p e r a t u r e of the tablets a r e required. a

The r e q u i r e d density for the c o r e s of fast r e a c t o r s (10.2-10.6 g / c m 3) is attained by baking at 1400~ however, this t e m p e r a t u r e is not sufficient for obtaining a homogeneous s t r u c t u r e . The incomplete f o r m a t i o n

1065

of the solid solution at this t e m p e r a t u r e is indicated by the phenomenon of r e s i d u e of PuO z during dissolving of the c o r e in nitric acid. The i n c r e a s e of the baking t e m p e r a t u r e to 1720~ reduces the amount of undissolved r e s i d u e to 3% (Fig. 1). On using c h e m i c a l l y coprecipitated p o w d e r s , the undissolved residue does no~ exceed 0.2%. F o r c o r e s of (U, tM)O z made f r o m powders obtained by the method of c h e m i c a l coprecipitation, it is c o n s i d e r a b l y s i m p l e to obtain a homogeneous s t r u c t u r e of the c o r e . A p r e l i m i n a r y t r e a t m e n t of the p o w d e r s is not required; the baking t e m p e r a t u r e in a r g o n m e d i u m with 7% hydrogen does not exceed 1500~ tn the e a s e of the u s e of m e c h a n i c a l m i x t u r e s of UO2 and PuO2, the homogeneous s t r u c t u r e of c o r e s is obtained with s u i t a b l e p r e p a r a t i o n of the p o w d e r s , in p a r t i c u l a r , with c a r e f u l mixing (Fig. 2). Materials of Shells f o r Fuel E l e m e n t s with U r a n i u m - - P l u t o n i u m Fuel. The m a t e r i a l of the f u e l - e l e m e n t shell of t e s t - i n d u s t r i a l r e a c t o r s 0Kh16N153B s t e e l in austenized s t a t e has been developod~ which r e p r e s e n t s a 7-solid solution hardened by molybdenum and niobium, besides carbon. The light e l e m e n t s in the s t e e l have been balanced in such a way that a complex hardening is achieved and the incidence of u n d e s i r a b l e phases (a, or, • is reduced to a m i n i m u m . F o r this p u r p o s e the content of nickel was i n c r e a s e d to 15%, the content of c h r o m i u m was limited to 16%, the molybdenum (3%) and niobium w e r e introduced into the steel. Depending on the t e m p e r a t u r e of austenization, its s t r u c t u r e contains different amounts of niobium c a r b o n i t r i d e s . R e sidual phases a r e evolved out with prolonged aging of the steel. During the inspection of i r r a d i a t e d s t e e l in t r a n s m i s s i o n , e l e c t r o n - m i c r o s c o p e dislocation loops of i n jected atoms a r e detected in its s t r u c t u r e besides the holes dotted by a l a r g e n u m b e r of fine extrusions of Nb(C, N) of 30 A. It is a s s u m e d that these niobium c a r b o r d t r i d e s block the growth and d i s p l a c e m e n t of the loops; they c e a s e to be sinks f o r the injected a t o m s , which facilitates the i n t e r a c t i o n of ~he t a t t e r with the holes and thus r e d u c e s swelling. The s t r u c t u r e of 0Khl6N15M3B s t e e l a f t e r i r r a d i a t i o n i n a BOR-60 r e a c t o r with a flux of 5.1022 n e u t r o n s / c m 2 is shown in Fig. 3a. The swelling c o m p r i s e s 0.3% [2]. The 0Khl6N15M3B s t e e l can e n s u r e the efficiency of the fuel e l e m e n t s with mixed oxide fuel up to a depletion of m o r e than 10% of the heavy a t o m s . However, a f t e r a c e r t a i n t i m e (at l a r g e i r r a d i a t i o n doses) all carbon, nitrogen, and niobium get evolved out of ~/-solid solution and Nb(C, N) p a r t i c l e s begin to grow. The s t r u c t u r e of 0Khl6N15M3B s t e e l a f t e r i r r a d i a t i o n by neutrons up to a flux of 7. t022 n e u t r o n s / e r a 2, when the swelling c o m p r i s e s 6.5%, is shown in Fig. 3b. The l a r g e p a r t i c l e s Nb (C, N) no longer r e t a r d swelling; furthermore, the presence of stresses around the large carbides activates sinks for the vacancies and, as shown by the electron microscope, pores form around the carbonitrides. This leads to an increase of the swelling. The tendency of the steel to high-temperature brittleness and interaction with the fission products of the field also increases. Therefore, a further improvement of the composition and the structural state of the steel was required for reactor steel; as a result, only a small radiation damage at the shells was obtained at a flux exceeding I. 1023 neutrons/cruZ(E7 0.i MeV). The Investigation of Irradiated Fuel Elements with Mixed Oxide Fuel. The efficiency of fuel elements of fast reactors made from mixed fuel is currently being investigated by radiation tests in the BOR-60 reactor. In the irradiated fuel elements the density of the tablets was varied (10.2-10.6 g/cm 3, i.e., 87-96% of the theoretical); the initial gap (0.i-0.25 ram) and the temperature of irradiation (620-680~ were also varied. The maximum linear load on the fuel element was 530 W/cm. The composition of the fuel was U0.85Pu0. ~5Oz. The oxygen coefficient was 1.97-1.98. Shells of 0KhI6NI5M3B steel did not lose hermeticity. The fuel elements reached depletion of 11% of heavy atoms at a flux of 7.1022 neutrons/cm 2 (E > 0.1 IvleV). The maximum change of the diameter of the shell was 1.5-4.2%. In order to estimate the contribution of radiational swelling of the steel and the inelastic deformation to the overall change of the diameter, the density of the samples cut from several segments of the shell was measured (Fig. 4). A comparison of the obtained results with the computed volume changes [from the ratio AV/V = 3(&d/d)] indicates that the inelastic deformation of the shell did not exceed 0.3% on the average, i.e., the contribution of the radiational creep caused by the pressure of the core to the swelling is relatively small. The gas release comprised 90%. The density of the fuel did not affec~ the gas release. In the upper par~ of the fuel element, where the temperature of the shell is the maximum, an interaction of the steel with the core is detected to a depth of 20-30 p with local damages up to 70 p. Corrosion interaction is caused by the effect of the fission products, especially cesium [3]. The shells still retained a certain margin of hardness and plasticity. The yield limit at 700~ is equal to 16 kgf/mm2; the relative elongation is about 1%. These results indicate that the fuel elements can operate up to deeper depletion. Fuel Cycle of Fast Reactors. The development of nuclear energetics with fast reactors substantially depends on the organization of production of fuel cycles, in particular, on the time required for processing of the reactive fuel and preparation of new- fuel from plutonium obtained from breeding.

1066

The amount of plutonium ensuring the operation of a single fast r e a c t o r can be d e t e r m i n e d in the following way:

where Gf is the total amount of plutonium in the fuel cycle of the reactor; G O is the plutonium loading in the active zone of the reactor; T o is the period of operation of the fuel in the reactor for a specified depletion; Th, Ttr , Trc, Tpr are, respectively, the time of holding, transportation, radiochemical treatment, and preparation of the fuel. The margin in the first reloading is 0.3 The efficiency of the fast reactor is usually characterized bythe time of doubling of the power excess breeding of plutonium. The doubling time can be defined by the formula

due to

Gf

T2 = 0.693 (BC--t) q'

(2)

where BC is the breeding coefficient of plutonium in the r e a c t o r and q is the amount of plutonium burned in the reactor per year. Equation (2) shows that the rate of introduction of atomic power plants with fast reactors due to plutonium breeding is inversely proportional to Gf; hence, it substantially depends on the time of the outer fuel cycle (the sum in square brackets in formula [i]). Investigations have shown that the holding of the fuel for i year before processing does not cause serious technical difficulties for the process of cleaning of the fissionable isotopes of the fission products and does not lead to complex problems in transportation to the radioehemical plant. If it is assumed that Gf must not be larger than 2G 0 then for T O = lo4y and Ttr = 1 month the time of radiochemical treatment of annual load of the reactor and also the time of preparation of the fuel elements for annual loading should not exceed i-1.5 months. In this case, after substituting the appropriate values into Eq. (i) we get Gf = (2.1-2.15) G 0. Thus, the capacity of the radioehemieal plant for the preparation of fuel elements must ensure reprocessing of the fuel from 8-10 atomic power stations with fast reactors, i.e., the treatment of the fuel must be organized centrally at a large capacity plant. Similar requirements

are imposed

also for a plant for preparing

heat-release

elements

from plutonium.

The increase of the holding time of the fuel up to 2-3 years before chemical treatment, which is tempting from the point of view of simplifying transportation and reprocessing of the fuel, leads to an increase of Gf~ (2.8-3.5) G 0. The rates of introduction of atomic power plants with fast reactors decreases by a factor ofi.31.6. The decrease of the time of holding of the fuel to 0.5 yr before chemical treatment lowers Gf to 1.8G 0 and increases the introduced capacity of atomic power plants with fast reactors by 15-17% with appreciable complications of the transportation operation and the process of radiochemical treatment. For an atomic po~er plant with a reactor and electrical capacity of 1600 MW and plutonium load in the zone of 3 tons (239pu + 2411>u) with BC = I.3, q = 1.2 tons and Th = 1 yr the time of doubling T 2 = 0.693Gf/(BC --l)q = 12 years. On shortening the time of holding to 0.5 yr the time of doubling decreases ing the holding time to 3 years it increases to 20 years.

to i0 years,

while on increas-

On the Use of Carbide and Nitride Uranium--Plutonium Fuel in Fast Reactors. One of the possible methods of improving the economic clmracteristics of power installations with fast reactors is to use such types of fuels as solid solutions of uranium and plutonium carbides and nitrides which have better thermophysical characteristics compared to the oxides. But the production of cores is somewhat more complex than in the case of oxides, since regulated highly clean inert atmosphere and high baking temperature of the cores (up to 2000~ are required [4]. The behavior of carbide--nitride fuel during irradiation differs from that of the oxide fuel: the structural changes occur more slowly and mainly in the central "hot" part of the samples; the gas release is appreciably smaller (~20%), cracking of the core is less pronounced, and the swelling increases. The mean volumetric swelling of the core made of (U, Pu)C comprises 2.3% at 1% depletion at a temperature of 1200~ atthe centerand* at depletion up to 7% of heavy atoms. An appreciable accumulation of gaseous fission products in carbide fuel leads to an increase of the core also in height up to 3% (for UC) for a linear power of 535 W/cm and depletion of 10% of the heavy atoms. The behavior of the fuel-element shells with cores of carbide and oxide fuels is on the whole similar. However, the contribution of the mechanical *Data missing from Russian Original - Publisher.

1067

interaction of the core with the shell to the diametral deformation swelling (up to 1.7%).

is noticeably larger compared

to neutron

A fast i n c r e a s e of swelling at a t e m p e r a t u r e higher than 1300~ r e s t r i c t s the specific energy r e l e a s e . An i n c r e a s e of the power output is possible if the fuel element is filled by sodium. However, in this c a s e the p r o c e s s of carbidization and softening of the steel shell gets a c c e l e r a t e d and uncertainties related to boiling and possible leakage of sodium appear. Computational investigations have shown [5] that fuel elements with carbide fuel can hold loads up to 750 W / c m and e n s u r e energy r e l e a s e up to 300 W/g. Investigations of nitride fuel have been c a r r i e d out to a l e s s e r extent and so f a r the amount of experimental data on the efficiency of such fuel elements is inadequate.

CONCLUSIONS The use of plutonium in the fuel cycle during complex utilization of thermal and fast reactors in nuclear energetics p e r m i t s solving the p r o b l e m of ensuring nuclear fuel f o r a long period. Oxide u r a n i u m - p l u t o n i u m fuel facilitates th~ development of technology of fast r e a c t o r s and so f a r it is considered as the basic type of fuel. At the s a m e t i m e , oxide fuel cannot e n s u r e the required rate of plutonium accumulation, in view of which the investigations of m o r e efficient fuel and constructional m a t e r i a l s become a p r e s s i n g problem. The use of uranium--plutonium oxide fuel in t h e r m a l r e a c t o r s r e q u i r e s i m p r o v e m e n t s in the construction of fuel elements and organization of l a r g e - s c a l e completely automatic production. E d i t o r s ' R e m a r k s . F o r the completeness of the discussion of the p r o b l e m it is, of c o u r s e , n e c e s s a r y to c o n s i d e r the possibility of using plutonium in fast and t h e r m a l r e a c t o r s as done by the authors. However, it should be kept in mind that by its n u c l e a r - p h y s i c a l p a r a m e t e r s p l u t o n i u m as a nuclear fuel is m o r e suitable for use in fast r e a c t o r s than in t h e r m a l r e a c t o r s . The use of plutonium in t h e r m a l r e a c t o r s can reduce the demands of natural u r a n i u m for the development of nuclear power in all by 10-15%, whereas its use in fast r e a c t o r s reduces the demand for uranium by a f a c t o r of 10. All this indicates the feasibility of using plutonium only in fast r e a c t o r s even if its accumulation is required over a certain period.

LITERATURE 1.

2. 3.

4.

5.

1068

CITED

V. V. Orlov, At. Energ., 36, No. 5, 341 (1974). N. P. Agapova et al., Problems of Atomic Science and Engineering, Series Fuel and Constructional Materials, No. 1 (4), TsNIIatominform, Moscow (1976), p. 3. M. Coquerelle, Trans. Am. Nucl. Soc., 1975. I. S. Golovnin et al., "Preparation and test of fuel elements from carbide fuel up to depletion of 3, 5, 7% of heavy atoms, " Paper presented at the winter session of American Nuclear Society, Washington (1974). I. S. Golovnin et al., At. Energ., 30, No. 2, 216 (1971).