Fuel assembly for thermal neutron type reactor

There is provided a fuel assembly for a thermal neutron type reactor in which fuel rods utilizing U-235 enriched uranium-oxide are bundled in a lattice arrangement, and a portion of the fuel rods in the lattice arrangement is substituted with MOX fuel rods provided with no U-235 enriched uranium-oxide fuel portion and no natural or depleted uranium-oxide blanket portion.

BACKGROUND ART 
The present invention relates to a fuel assembly adapted for a thermal 
neutron type reactor and more particularly to a fuel assembly utilizing 
fuel rods enriched in plutonium. 
In view of effective usage of resource and energy security, there is a 
schedule for the utilization of plutonium recovered through reprocessing 
of spent fuel as a fuel in a thermal neutron reactor. 
The plutonium radiates .alpha.-rays having high radiation intensity and it 
is hence necessary to prevent a human body from being internally exposed 
and also radiates neutrons and .gamma.-rays through decay and spontaneous 
fission. For this reason, production or fabrication of the fuel including 
the plutonium should be performed in a sealed environment in comparison 
with uranium fuel. In addition, many considerations must be paid for 
equipment and manufacturing processes. For example, multiple shielding 
equipment is required and strict attention should be paid for its 
decontamination and maintenance. Accordingly, it is extremely 
disadvantageous from economical and other view points to manufacture many 
kinds of fuel pellets and fuel rods containing the plutonium in different 
concentrations. 
From another view point, since severe conditions are placed on the fuel 
rods containing the plutonium with respect to the conveyance, measurement 
control and criticality control, it is desired to reduce the number of 
fuel rods containing the plutonium by using a large containing ratio of 
plutonium in one fuel rod. 
From the above view points, in a fuel assembly in which plutonium of a 
predetermined amount obtained through the reprocessing of spent fuel is 
utilized for a thermal neutron reactor, it is advantageous to substitute 
fuel rods each having high enrichment, in such a fuel assembly utilizing 
enriched uranium as shown in FIG. 7, with fuel rods containing the 
plutonium such as uranium-plutonium mixed-oxide fuel (MOX). 
FIG. 7A shows a fuel arrangement in the radial direction and FIG. 7B shows 
a fuel arrangement in the axial direction. In these figures, reference 
numeral 1 denotes a channel box, 2 denotes a fuel rod, a symbol Ui (i=1-4) 
represents a uranium fuel, G is a fuel rod containing a burnable poison 
and W is a water rod. 
In the substitution of the highly enriched uranium fuel rods in the uranium 
fuel rod assembly with the MOX fuel rods, the enrichment of the 
fissionable plutonium is experimentally set such that the reactivity 
characteristics as the plutonium fuel assembly and the peaking factors in 
the radial and axial directions becomes approximately the same as those of 
the uranium fuel assembly, but as a result, the following relationship 
will be established approximately. 
Namely, in the case where the fuel rod having enrichment ei of the uranium 
fuel assembly is substituted with the MOX fuel rod having the enrichment 
Pi, supposing that the concentration of U-235 of the uranium of the MOX 
fuel is eB, the following equation (1) will be established. 
##EQU1## 
I: Numbers of the fuel rods to be substituted. 
In this equation Q becomes 1.2 to 1.5. 
In another case where the fuel rod having high enrichment of the uranium 
fuel assembly is substituted with the MOX fuel rod, there is a case in 
which blanket portions having low concentration of U-235 are arranged to 
upper and lower ends of the uranium fuel rod. However, in the MOX fuel 
rods, there is a possibility of increasing the number of MOX fuel rods 
when a predetermined amount of the recovered plutonium is treated with the 
MOX fuel rod by providing the blanket portions having a lower U-235 
concentration than natural uranium, thus being disadvantageous from an 
economical view. 
In the thermal neutron reactor, when it is required to mix the plutonium 
with the uranium fuel and then to cause a fission reaction, nuclides other 
than U-235 for causing the fission reaction with respect to the thermal 
neutrons are plutonium isotopes of Pu-239 and Pu-241. The plutonium 
further includes Pu-240 absorbing the thermal neutrons and minute amount 
of Pu-238 and Pu-242. 
The plutonium isotope Pu-241 is subjected to .beta.-decay with a relatively 
short half life (14.7 years) and decays to Am-241 as a neutron absorbing 
nuclide. During the cooling period of the spent fuel, Pu-238 and Pu-240 
are transformed by .alpha.-decay of Cm-242 and Cm-244, respectively, but 
the transformed amounts thereof are small and their influence on the 
characteristics of the MOX fuel will be neglected. 
The amount of fissionable isotope of plutonium to be enriched in the MOX 
fuel rod for a thermal neutron reactor is restricted to the amount 
necessary for keeping a chain reaction for a predetermined period and only 
the amount determined by design is mixed. Namely, U-235, Pu-239 and Pu-241 
must be mixed so as to take a value designed for achieving a predetermined 
reactivity. 
However, the plutonium is recovered by the reprocessing of the spent fuel 
and, accordingly, the isotope composition thereof differs in accordance 
with initial enrichment, burnup degree and cooling period of the spent 
fuel. FIG. 8 represents one example of the change of plutonium isotope in 
case of uranium fuel burnup. 
Accordingly, the plutonium recovered after reprocessing through mixing of 
various kinds of spent O.sub.2 fuels is considered to have a certain 
isotopic composition due to reprocessing. 
In general, the fuel design is determined so that the enrichment of the 
fissionable substance of the fuel rod has a predetermined reactivity 
before the recovery of the plutonium. However, it is complicated and 
disadvantageous to redesign the enrichment of the fissionable substance 
every time when the composition of the fissionable nuclide in the actually 
recovered plutonium differs in order to maintain the predetermined 
reactivity. 
Further, since Am-241 is removed as an impurity at the time of the recovery 
of the plutonium during the reprocessing of the spent fuel, it is not 
necessary to consider accumulation thereof during the spent fuel cooling 
period. However, the Am-241 is accumulated in a period from the recovery 
of the plutonium to its loading into a core as a manufactured MOX fuel, so 
that it becomes necessary to design the enrichment in consideration of its 
influence on the core characteristics. 
For the design of the enrichment of the MOX fuel, the use of the mixture of 
the recovered plutonium is assumed as described above, but the actual 
mixing ratio and the amount of Pu-241 to be transformed to Am-241 through 
the .beta.-decay during the cooling period and a time before its loading 
into the core after the reprocessing, are not clear. For this reason, in 
the conventional technology, the design has been made by tentatively 
assuming the containing rate of the sum Puf of the fissionable plutonium 
isotope Pu-239 and Pu-241 contained in the recovered plutonium by 
considering the initial enrichment, burnup degree, cooling period, and 
reprocessing amount, and the time after reprocessing before loading into 
the core as a fuel assembly after the transformation to PuO.sub.2 and 
manufacturing as the MOX fuel through mixing with UO.sub.2. 
That is, at a time when the MOX fuel using the recovered plutonium after 
the reprocessing of the spent fuel in a boiling water reactor BWR, on the 
assumption that plutonium would be obtained having a plutonium containing 
ratio F of from about 80%, which is recovered after reprocessing spent 
fuel having a low initial enrichment and low burnup degree such as 
initially loaded fuel, to about 60%, which is recovered after the 
reprocessing of spent fuel having a high initial enrichment and high 
burnup degree such as reloaded fuel; the larger containment rate of about 
80% is sought. This is because the reactivity becomes high in the case of 
the high containing rate F, that is, less amount of Pu-240 or Pu-242 as 
the thermal neutron absorbing nuclide, and accordingly, the margin with 
respect to the thermally limited value in the operational characteristic 
at loading in the core of the MOX fuel becomes small. According to the 
confirmation of the margin in the design on the assumption of such 
plutonium containing rate F, in the actual plutonium usage, in case the 
plutonium containing rate F could be made to be larger than the assumed 
value in the design, the margin with respect to the thermally limited 
value could be made large. 
Further, with respect to the decay of Pu-241 and its daughter of Am-241 
which is accumulated during the time from manufacturing the recovered 
plutonium into the MOX fuel and loading into the core, a shorter 
arrangement such as about one year has been made in consideration of an 
actually usable period. This is also based on the consideration that 
Pu-241 is contained in a higher amount in the less usable period and a 
higher margin can be ensured in view of its reactivity. 
As described above, in the design of the MOX fuel, the enrichment should be 
set by preliminary assumption of the containing rate F with a certain 
composition in the initial design in spite of the fact that the 
composition of the plutonium and, in particular, the containing rate F of 
fissionable plutonium, to be actually obtained and used are not clear, and 
accordingly, in the initial design, the containing rate F should be set to 
a considerably large value with respect to the plutonium to be actually 
used. 
After the MOX fuel has been designed and the margin having an optimum 
operational characteristic has been confirmed, the plutonium to be 
actually used is obtained. In a case where the containing rate F of the 
plutonium obtained is different from that of the assumption at the design 
time, the mixing amount of the plutonium is regulated so that the amount 
of fissionable nuclides .sup.235 U+.sup.239 Pu+.sup.241 Pu are to 
coincide. In this method, however, the reactivity is increased or 
decreased largely in accordance with the increasing or decreasing of the 
amounts of Pu-240 and Pu-242. In the case of the amount of PU-240 and 
Pu-242 being larger than that in the design, less reaction may be caused 
and power may be hence reduced. From the viewpoint of ensuring the design 
margin, since safety is maintained in the case of less reaction than in 
the case of excessive reaction, the design is made as a countermeasure so 
that such isotopes as Pu-240 have the containing rate less than the 
assumed plutonium composition and, in the case where the reactivity 
shortage is actually caused, an amount of fuel to be exchanged is 
increased or the running period is made short. However, such 
countermeasures result in the change of the fuel amount to be required or 
disadvantages in economy. 
In the BWR as one example of the thermal neutron reactor, as shown in FIG. 
9, a number of fuel assemblies are allocated within a channel box to 
constitute a core. Referring to FIG. 9, reference numerals 3, 4, 5, 6 and 
7 denote a fuel assembly, a local power range monitoring system LPRM, an 
intermediate power range monitoring system IRM, a source range monitoring 
system SRM and a control rod, respectively. 
In a space between the adjacent fuel assemblies 3, there is located a water 
gap area, having a constant width, for arranging a cross-shaped control 
rod, i.e. control blade, or instrumentation tube. 
Coolant in the channel box constitutes two layer flows of water and steam 
during the running of the core, but in the water gap area, the coolant is 
not directly heated by the fuel rod, thus not generating steam. For this 
reason, the atomic density of hydrogen in the water gap area is large and 
the radical distribution of the thermal neutrons in the horizontal 
cross-section of the fuel assembly 3 of the BWR is made large in the 
peripheral portion of the fuel assembly. In order to make a power peaking 
factor small in an inner radial direction of the fuel assembly, it is 
adapted to arrange fuel rods having low enrichment to the peripheral 
portions of the fuel assembly as shown in FIGS. 7A and 7B. 
In a case where the fuel rods containing plutonium are used, in order to 
make a power peaking factor small in the radial and axial directions of 
the fuel assembly, it is necessary to adjust the enrichment and the 
distribution of the uranium fuel rods and the density and the distribution 
of a burnable poison as well as their arrangement in the fuel assembly. 
In general, the isotope of the plutonium recovered by the reprocessing is 
different, as described hereinbefore, in initial enrichment, burnup 
hysteresis, burnup degree, cooling period, etc. of the spent fuel 
reprocessed. However, when MOX fuel manufactured by mixing such plutonium 
with uranium is irradiated by thermal neutrons in the thermal neutron 
reactor, the isotopes of uranium and plutonium are transformed as follows. 
Namely, neutrons of the uranium fuel are absorbed by U-238 in the thermal 
neutron reactor and transformed into Pu-239. The Pu-239 is fissioned by 
the absorption of thermal neutrons, but a portion thereof is transformed 
into Pu-240, which is then transformed into Pu-241 by the absorption of 
neutrons. The Pu-241 is a fissionable nuclide, but a portion thereof is 
further transformed into Pu-242 by further absorption of neutrons. The 
Pu-241 is then transformed into Cm-242 by the absorption of neutrons. As a 
consequence of such reactions of the plutonium isotopes, the fissionable 
substance contained in the fuel rod is less reduced by the fission and the 
lowering of the reactivity does not progress due to the burnup of the fuel 
in comparison with the uranium fuel. Therefore, the power peaking of the 
fuel rod containing the plutonium in the radial direction of the fuel 
assembly has a tendency of being made large during the burnup in 
comparison with the fuel rod containing the uranium. 
Further, since the composition of the plutonium isotope recovered after the 
reprocessing cannot be specified, it is necessary to either not utilize 
the rod in a case where the composition of the plutonium actually obtained 
includes a larger amount than the Puf amount assumed in the design, or to 
mix the recovered plutonium with other plutonium containing a lesser 
amount of Puf. In a conventional technique, there is no clear standard for 
judgement with respect to a mixing ratio and compositions in such mixing 
of the plutonium isotopes, and accordingly, such mixing has been performed 
case by case on the basis of experimental results. 
As described above, in the conventional technology, when plutonium is 
mixed, there is no clear standard for judgement as to the mixing ratio and 
the composition to be obtained, so that the reactivity of the MOX fuel 
manufactured using the plutonium depends largely on the composition of the 
obtained plutonium, and the scattering of the reactivity is observed from 
a view point of interchangeability with the uranium, thus including 
disadvantageous points relating to the limit of the power peaking in core 
operation and in fuel economy. 
The present invention was conceived in view of the above defects and 
disadvantages and aims to provide a fuel assembly for a thermal neutron 
type reactor capable of reducing the kind and number of fuel rods 
containing plutonium and increasing the margin with respect to the thermal 
limit during the running of the reactor. 
Another object of the present invention is to provide a fuel assembly for a 
thermal neutron type reactor capable of easily setting the fission 
reaction effects of the fuel containing the plutonium and easily 
performing correction for maintaining its characteristics. 
A further object of the present invention is to provide a fuel assembly for 
a thermal neutron reactor capable of preventing an excessive increase of 
power peaking factor in the radial and axial directions of the fuel 
assembly containing the plutonium. 
DISCLOSURE OF THE INVENTION 
The first embodiment of this invention, for achieving the above objects, is 
characterized in that a portion or the whole of the fuel rods of a fuel 
assembly for a thermal neutron type reactor in which fuel rods utilizing 
enriched uranium are bundled in lattice arrangement is substituted with 
fuel rods containing plutonium and the fuel rods containing the plutonium 
are provided with no uranium fuel portion and no uranium blanket portion. 
The second embodiment of this invention, for achieving the above objects, 
provides a fuel assembly for a thermal neutron reactor in which fuel rods 
utilizing enriched uranium are bundled in lattice arrangement, which is 
characterized in that a portion of the fuel rods in the lattice 
arrangement is substituted with fuel rods containing plutonium, 
concentrations of the plutonium are set in accordance with the containing 
rate of Pu-240, an isotope of the plutonium, and with respect to plutonium 
having a composition of different plutonium isotopes, predetermined 
concentrations are obtained by deciding a mixing ratio in accordance with 
the difference in the amount of Pu-240 to thereby obtain a predetermined 
plutonium isotope composition. 
The third embodiment of this invention, for achieving the above objects, 
provides a thermal neutron type reactor in which fuel rods utilizing 
enriched uranium are bundled in lattice arrangement, which is 
characterized in that a portion or the whole of the fuel rods in the 
lattice arrangement is substituted with fuel rods containing plutonium, 
concentrations of the plutonium are set in accordance with the containing 
rate of Pu-240, an isotope of the plutonium, and with respect to plutonium 
having a composition of different plutonium isotope, predetermined 
concentrations are obtained by deciding a mixing ratio in accordance with 
the difference in the amount of Pu-240 to thereby obtain a predetermined 
plutonium isotope composition. 
The fourth embodiment of this invention, for achieving the above objects, 
provides a fuel assembly for a thermal neutron type reactor in which fuel 
rods utilizing enriched uranium are bundled in lattice arrangement, which 
is characterized in that a portion of the fuel rods in the lattice 
arrangement is substituted with fuel rods containing plutonium and 
concentration of the plutonium are corrected in accordance with the 
composition of the plutonium, cooling period of spent fuel and period 
until the time of mixing the plutonium with the uranium to manufacture the 
fuel and loading into a core after reprocessing treatment. 
The fifth embodiment of this invention, for achieving the above objects, 
provides a thermal neutron type reactor in which fuel rods utilizing 
enriched uranium are bundled in lattice arrangement, which is 
characterized in that a portion of the fuel rods in the lattice 
arrangement is substituted with fuel rods containing plutonium and the 
concentration Pi of the plutonium is set by the following equation 
including an concentrations ei of the uranium fuel 
EQU Pi=(ei-eB).multidot.Q 
EQU Q.ltoreq.1.2 
eB: U-235 concentration contained in uranium of a fuel mixed with plutonium 
According to the fuel assembly for a thermal neutron type reactor of the 
first embodiment, the uranium fuel portion and the uranium blanket portion 
are not provided. For this arrangement, the plutonium amount contained in 
one fuel rod increases and the number of the fuel rods manufactured from a 
predetermined amount of plutonium can be reduced. Furthermore, the 
effective heat generation length of the fuel rod containing plutonium is 
made long and the maximum linear power density is lowered, whereby the 
margin with respect to the thermal limit value during running core can be 
increased. 
Further, according to the fuel assembly for a thermal neutron type reactor 
of the second embodiment, the concentration of the plutonium of the fuel 
rod containing the plutonium is set in accordance with the containing rate 
of Pu-240. The relationship between the sum Puf (Puf=.sup.239 Pu+.sup.241 
Pu) of the fissionable nuclides in plutonium isotopes immediately after 
the irradiation of spent fuel and the Pu-240 is represented by a simple 
descending curve which is substantially not dependent on the spent fuel 
and its burnup degree. For this reason, by utilizing this simple 
descending relation, the reactivity change caused by the difference in the 
plutonium isotope composition can be easily and reasonably compensated 
for. 
Furthermore, according to the fuel assembly for a thermal neutron type 
reactor of the third embodiment, with respect to the plutonium having the 
different isotope composition, the mixing ratio is decided in accordance 
with the difference of the Pu-240 amount to thereby obtain the 
predetermined plutonium isotope composition. Accordingly, even in a case 
where the amount of Pu-240 assumed at the design time differs from the 
amount of the actually obtained Pu-240, the concentration correction can 
be done by making the simple descending curve approximate a straight line. 
Still furthermore, according to the fuel assembly for a thermal neutron 
type reactor of the fourth embodiment, the concentration of the fuel rod 
containing the plutonium is corrected in accordance with the composition 
of the plutonium, the cooling period of the spent fuel and the period 
until the time of mixing the plutonium and the uranium, MOX fuel 
manufacturing and then loading it into the core, thus being capable of 
accurately setting the concentration. 
According to the fuel assembly for a thermal neutron reactor of the fifth 
embodiment, the concentration of the plutonium of the fuel rod containing 
the plutonium is set by the relationship of a predetermined equation 
substantially including the enrichment of the uranium fuel. For this 
reason, the power peaking factor of the fuel rod containing the plutonium 
can be lowered relatively to that of the uranium fuel rod.

BEST MODES FOR CARRYING OUT THE INVENTION 
Hereinbelow, a first embodiment of the present invention will be described 
with reference to the accompanying drawings. 
FIGS. 1A and 1B represent an MOX fuel assembly for a BWR as one example of 
a fuel assembly for a thermal neutron type reactor according to the 
present invention, in which FIG. 1A shows a fuel arrangement in the radial 
direction of the fuel assembly and FIG. 1B shows a fuel arrangement in the 
axial direction thereof. 
In FIGS. 1A and 1B, reference numeral 1 denotes a channel box and 2 is a 
fuel rod, and a symbol Ui (i=1-4) denotes uranium fuel, P1 is an MOX fuel 
containing plutonium, G is a fuel rod containing a burnable poison and W 
is a water rod. 
In the present embodiment, the fuel assembly has a structure in which the 
eighteen highly enriched fuel rods disposed at the central portion of a 
conventional uranium fuel assembly such as shown in FIGS. 7A and 7B are 
substituted with MOX fuel rods. 
Further, the concentrations of the fissionable plutonium (Pu), the U-235 
enrichments of and the poison density of the fuel rod are, as follows. 
Concentration of Fissionable Pu 
P1: 4.5% 
Enrichment of U-235 
e1: 4.0% 
e2: 3.0% 
e3: 3.8% 
e4: 2.8% 
e5: 3.5% 
e6: 2.3% 
e7: 2.6% 
e8: 1.8% 
Enrichment of U-235 of Fuel Rod Containing Burnable Poison 
eg1: 4.1% 
eg2: 4.9% 
Poison Density of Fuel Rod Containing Burnable Poison 
g1: 3.5% 
g2: 4.5% 
Each of these eighteen MOX fuel rods has a structure corresponding to that 
of FIGS. 7A and 7B in which the uranium fuel rods having an enrichment 
higher than that of fuel rods disposed to the peripheral portion of the 
fuel assembly are substituted with the MOX fuel rods. The concentration of 
the plutonium is made high by substituting the fuel rods disposed to the 
peripheral portion with the MOX fuel rods, and accordingly, the numbers of 
fuel pellets and fuel rods can be reduced at a time of fabricating a 
predetermined amount of the recovered plutonium to the MOX fuel. 
FIGS. 2A and 2B represent a second embodiment of the present invention, in 
which forty fuel rods having high enrichment disposed at the central 
portion of the uranium fuel assembly and a part of the peripheral portion 
thereof are substituted with the MOX fuel rods. In FIGS. 2A and 2B, a 
symbol Uj (j=1-3) denotes a uranium fuel rod and Pi (i=1-3) is an MOX fuel 
rod containing plutonium. The concentrations of the plutonium and the 
enrichments of uranium U-235 are as follows. 
Concentration of Fissionable Pu 
P1: 6.2% 
P2: 5.0% 
P3: 3.6% 
Enrichment of U-235 
e1: 4.0% 
e2: 3.0% 
e3: 3.3% 
e4: 2.3% 
e5: 2.2% 
e6: 1.8% 
Enrichment of U-235 of Fuel Rod Containing Burnable Poison 
eg1: 4.1% 
eg2: 4.9% 
Poison Density of Fuel Rod Containing Burnable Poison 
g1: 1.5% 
g2: 2.5% 
In the present embodiment, twenty uranium fuel rods not substituted with 
the MOX fuel rods have low enrichment less than half of the enrichement of 
the centrally arranged fuel rods. For this arrangement, the amount of the 
MOX fuel rods to be manufactured having low plutonium concentration is 
reduced by substituting the fuel rods disposed to these portions with the 
uranium fuel rods, and accordingly, the number of MOX fuel rods 
manufactured from the predetermined amount of the recovered plutonium can 
be effectively reduced. 
FIGS. 3A and 3B represent a third embodiment of the present invention, in 
which MOX fuel rods each having no natural uranium blanket portion at its 
upper and lower ends. The concentrations of the plutonium and the 
enrichments of uranium-235 are as follows. 
Concentration of Fissionable Pu 
P1: 6.2% 
P2: 5.0% 
P3: 3.6% 
Enrichment of U-235 
e1: 4.2% 
e2: 3.9% 
e3: 3.4% 
e4: 3.9% 
e5: 3.4% 
e6: 2.5% 
e7: 3.4% 
Enrichment of U-235 of Fuel Rod Containing Burnable Poison 
eg1: 2.8% 
eg2: 3.0% 
eg3: 2.8% 
Poison Density of Fuel Rod Containing Burnable Poison 
g1: 3.5% 
g2: 4.5% 
g3: 2.5% 
In the present embodiment, the natural uranium blanket portion is not 
provided for the MOX fuel rod, so that the number of the MOX fuel rods 
manufactured from the predetermined amount of recovered plutonium can be 
prevented from increasing. Furthermore, since the blanket portion is 
constructed to be an effective portion for heat generation as fuel 
portion, the maximum linear power density can be lowered and the margin to 
the thermal limit during core running period can thus be increased. 
FIG. 4 represents a fourth embodiment of the present invention adapted for 
a fuel assembly including fuel rods in 9-row and 9-line arrangement 
(9.times.9 fuels). 
In this embodiment, the concentrations of the plutonium and the enrichments 
of uranium-235 are as follows. 
Concentration of Fissionable Pu 
P1: 6.2% 
P2: 5.0% 
P3: 3.6% 
Enrichment of U-235 
U1: 3.5% 
U2: 2.8% 
U3: 2.0% 
Enrichment of U-235 of Fuel Rod Containing Burnable Poison 
upper portion: 4.9% 
lower portion: 4.1% 
eg3: 2.8% 
Poison Density of Fuel Rod Containing Burnable Poison 
upper portion: 2.5% 
lower portion: 1.5% 
In the above respective embodiments, the transformation of the composition 
of the plutonium isotope during the irradiation in a thermal reactor is 
mainly carried out through the following chain reaction by the absorption 
of neutrons. 
EQU U-238.fwdarw.Pu-239.fwdarw.Pu-240.fwdarw.Pu-241.fwdarw.Pu-242(2) 
A portion of each of Pu-239 and Pu-241 is burned up through the fission. 
A following equation shows a sum of the fissionable nuclides in the 
plutonium isotope immediately after the irradiation of the spent fuel. 
EQU Puf=.sup.239 Pu+.sup.241 Pu (3) 
The relationship between Puf and Pu-240 is represented in FIG. 5. 
That is, the relationship between the Pu-240 and the Puf is represented by 
a simple descending curve substantially not dependent on the spent fuel 
and its burnup degree. Accordingly, the composition of the plutonium 
isotope after the mixing can be obtained by the amount of the Pu-240 and 
the mixing ratio even if the plutonium has a different composition due to 
the reprocessing batch by applying this simple descending curve to an 
approximation of a straight line. 
According to the present invention, the setting of the composition of the 
recovered plutonium after the mixing can be easily and reasonably made by 
utilizing the simple descending relationship between the Pu-240 and the 
Puf. 
In a case where the composition of the recovered plutonium isotope is 
different from the composition tentatively assumed at the time of MOX fuel 
design, the reactivity change caused by the difference of the plutonium 
isotope composition of the MOX fuel can be easily and reasonably 
compensated for by utilizing that simple descending relationship. 
Now, supposing that the simple descending curve of Puf at the containing 
ratio of Pu-240 of the isotopic composition of the recovered plutonium set 
at the stage of MOX fuel design crosses with a straight line having the 
same inclination as that of a tangential line on that simple descending 
curve at points fA and fB as shown in FIG. 6, in which, the intersections 
fA and fB of the curve and the straight line are set so that the Pu-240 
containing ratio A for one plutonium batch obtained by reprocessing spent 
fuel and the Pu-240 containing ratio B for another plutonium batch 
obtained by reprocessing other spent UO.sub.2 fuels, respectively. Pu-240 
of the mixture of the plutonium A and the plutonium B has a point C in 
FIG. 6 at which the weighted average is made with the respective weights, 
and the Puf is positioned on a point fC', corresponding to the point C, on 
the line connecting the points fA and fB. Accordingly, by mixing the 
plutonium so that the amount C of the Pu-240 coincides with the amount D 
of the Pu-240 of the composition of the plutonium isotope assumed at the 
MOX fuel assembly design time, the amount fC' of the Puf at that time is 
only slightly lower than the amount fd assumed at the design time, and 
hence, characteristics in reactivity of the MOX fuel assembly are 
substantially not changed from evaluation at the design time or, even if 
changed only a slight increase of the operational margin will be caused. 
Further, at the time of the MOX fuel assembly design, there is a tendency 
of often setting the amount of Puf to a relatively smaller value from the 
reason that the composition of the plutonium isotope actually recovered is 
not clear. In such case, the amount C of Pu-240 of the actually obtained 
plutonium becomes larger than the amount D of Pu-240 tentatively set at 
the design time, and accordingly, the characteristics in the reactivity of 
the MOX fuel assembly become smaller than that of the design time by an 
amount corresponding to .DELTA.fd as shown in FIG. 6A. This is 
disadvantageous for the reactivity characteristics of the MOX fuel 
assembly. In order to make equivalent in the reactivity the plutonium of 
such composition to the MOX fuel utilizing the plutonium having the 
composition at the design time, the design value of the concentrations of 
the plutonium of the MOX fuel will be corrected to make the same high, 
whereby the amount of the MOX fuel to be manufactured can be reduced, thus 
being advantageous in fuel economy. 
Furthermore, according to the present invention, the reactivity change due 
to the difference between the amount of Pu-240 actually obtained and the 
amount of Pu-240 assumed at the design time can be compensated for by a 
reasonable method. 
That is, now supposing that the Pu-240 amount of the obtained plutonium, 
with respect to the concentrations Pi of the MOX fuel rod, is the value of 
C in spite of the design with the Pu-240 amount of D of the composition of 
the plutonium isotope. In such case, the amount of the Puf is less by the 
amount of fd-fc=.DELTA.fd, and an amount larger by C-D of Pu-240 being the 
neutron absorbing nuclide from the assumed amount D at the design time is 
mixed with the MOX fuel rod having the concentration Pi. However, 
according to the present invention, in which the concentration Pi could be 
corrected as shown in the following equation in accordance with the 
difference of the Pu-240 amounts by making approximate, to the straight 
line as described hereinbefore, the fissionable substance of the 
composition of the plutonium isotope with the amounts of C and D of 
Pu-240. That is, the correction amount is expressed as, by using the 
symbols in FIG. 6A, 
EQU {(fA-fB)/(A-B).multidot.(D-C)+fC'}.times.Pi/fC'=fd'/fC'.times.Pi(4) 
Thus, the enrichement Pi is multiplied by fd'/fC' times. 
According to this correction, the Puf ratio of the actually obtained 
plutonium becomes fC+.DELTA.fC'=fd' slightly smaller than the Puf ratio of 
the plutonium at the design time. Therefore, the Puf ratio becomes 
substantially equivalent to that at the design time in the view point of 
the reactivity, and in the view point of the operational margin, the 
margin does not unnecessarily increase, thus providing a reasonable 
correction amount according to the present invention. Further, in a case 
where the Pu-240 amount of the obtained plutonium differs from that of the 
plutonium set at the design time, the inclination of the above-mentioned 
straight line can be made as shown in FIG. 6B in a manner such that the 
inclination has a range more than an inclination of the tangential line to 
the above-mentioned curve when the Pu-240 containing ratio of the 
recovered plutonium is less than that set tentatively at the design time 
and that the inclination has a value less than that of the inclination of 
the tangential line when the Pu-240 containing ratio of the recovered 
plutonium is larger than that set tentatively at the design time. 
Namely, when the Pu-240 of the obtained plutonium is C1 which is smaller 
than the Pu-240 containing ratio D set tentatively at the design time, the 
fC'l at the value of C1 of the line fA'-fB having an inclination less than 
the inclination of the above-mentioned tangential line at the point fd and 
the fd' at the value of D have relationship of fd'/fc'l.ltoreq.fd/fC1. 
Accordingly, the correction amount can be made small and hence an 
excessive correction will be avoided by multiplying the Pi by fd'/fC'l 
times other than fd/fcl times. In a case where the containing ratio of the 
Pu-240 of the obtained plutonium has a value C2 which is larger than the 
value D, fC'2 at the ratio C2 and fd' at the ratio D of a line fA-fB' 
having an inclination not exceeding the inclination of the tangential line 
at fd has a relationship of fd'/fC'2.ltoreq.fd/fC2, an excessive 
correction can be obviated by multiplying the Pi by fd'/fC'2 times other 
than fd'/fc'2 times. 
Further, supposing that the amount of Pu-241 at a time just after discharge 
of the spent fuel is .sup.241 Pto, the amount of Pu-241 and the .sup.241 
Pto+.DELTA.t at the cooling period of (to+.DELTA.t) are expressed as 
follows. 
EQU .sup.241 Pto+.DELTA.t=.sup.241 Pto(1-e.sup.-.lambda..DELTA.t)(5) 
.lambda.=0.693/T1/2 T1/2: half life (14.7 year) 
According to the present invention, the reduction of the Pu amount from the 
above equation forms a method of correcting the reactivity loss due to 
Pu-241 decay, thereby easily setting the reactivity effects of the MOX 
fuel by reasonably carrying out the plutonium concentration design of the 
MOX fuel. 
Namely, for the relation of the Puf with respect to the aforementioned 
Pu-240, the simple descending curve is established in no consideration of 
the cooling period, so that, when the concentration of the plutonium in 
MOX fuel is to be set, the concentration of plutonium can be made high so 
as to compensate for the reduced amount during the cooling period of the 
Pu-241 amount from the fissionable Pu ratio Cf of the obtained plutonium 
in view of the above relation. 
Further, in a case where the recovered plutonium after the reprocessing is 
treated with an oxide, the Am-241 produced through the .beta.-decay of the 
Pu-241 is accumulated. The Am-241 absorbs neutrons and a portion thereof 
is transformed to fissionable Am-242, so that the reactivity is recovered 
to some extent during the irradiation by thermal neutron in the thermal 
reactor. 
An amount At1 of the Am-241 at .DELTA.t1 after the reprocessing will be 
expressed as follows on the assumption of the Pu-241 amount after the 
reprocessing to .sup.241 Pu (t1=0). 
EQU At1=.sup.241 Pu(t1=0).multidot.(1-e.sup.-.lambda..DELTA.tl)(6) 
According to the present invention, as a method of correcting the 
reactivity loss due to the accumulation of the Am-241, an accumulated 
amount of Am-241 is calculated from the above equation and the setting of 
the concentration of the MOX fuel can be reasonably performed. Namely, at 
the setting of the concentration of the MOX fuel, the accumulated amount 
of the Am-241 is assumed from the above equation with respect to a period 
between the assumed manufacturing to the MOX fuel and the loading into the 
core, to thereby make the plutonium concentration high so as to correct 
the reactivity loss effect. 
Furthermore, in a fuel assembly in which a conventional enriched uranium is 
used, when a portion of the fuel rods using the enriched uranium is 
substituted with the MOX fuel rods, the following procedure will be 
carried out at the setting time of the plutonium concentration of the MOX 
fuel rods in order to prevent the excessive increasing of the power 
peaking factor in the radial and axial directions of the MOX fuel assembly 
caused by the difference with respect to the reduction amount of the 
fissionable substance in the uranium fuel rod due to the production of the 
fissionable substance due to the nuclear transformation during the 
irradiation of the plutonium in the MOX fuel rod. 
That is, in the conventional technology, as mentioned above, when the 
uranium fuel rods having the enrichment ei are substituted with the MOX 
fuel rods, the concentration P1 of the fissionable substance was expressed 
as follows. 
##EQU2## 
However, according to the present invention, the power peaking factor of 
the MOX fuel rod in the MOX fuel assembly can be relatively lowered with 
respect to that of the uranium fuel rod. In the above equation, symbol eB 
represents a density of U-235 in the MOX fuel rod and I represents the 
number of the fuel rods. 
Still furthermore, according to the present invention, in order to prevent 
the excessive increasing of the axial power peaking factor of the MOX fuel 
assembly, provision is made in related to the enrichment distribution or 
the density distribution of the burnable poison in the axial direction of 
the uranium fuel rod or fuel rod containing the burnable poison in the MOX 
fuel assembly. 
Namely, in the conventional uranium fuel assembly, the enrichment 
distributions in the axial direction are provided for a number of fuel 
rods. Accordingly, if these fuel rods are substituted with the MOX fuel 
rods, the arrangement of the enrichment distributions of the fissionable 
plutonium, results in increasing the kinds of MOX fuels. However, the 
enrichment distribution or the density distribution of the uranium 
fuel-rods or fuel rods containing the burnable poison, which are not 
substituted with the MOX fuel rods, are arranged to thereby constitute the 
MOX fuel assembly and the excessive increasing of a power peaking factor 
in the axial direction can be prevented. 
Further, in the above-mentioned fuel assembly, a MOX fuel manufacturing 
line can be reduced by using uranium fuel without plutonium, thus reducing 
the manufacturing cost. 
Furthermore, according to the present invention, fuel containing plutonium 
and fuel containing no plutonium can be commonly used as the fuel rods 
containing the burnable poison. In the view point of reducing the 
manufacturing cost of the plutonium fuel, it is advantageous to use the 
uranium fuel as the base. Thus, the plutonium loading amount is reduced 
per one fuel assembly, and accordingly, in a case where a large loading 
amount of the plutonium is heavily weighed, the burnable poison will be 
mixed into the fuel containing the plutonium. 
The present invention achieves the axial distribution of the fissionable 
substance (enrichment) or burnable poison amount for the fuel of the BWR 
by the providing of the fuel rods containing plutonium and the uranium 
fuel rods containing the burnable poison. 
Hereinbelow, the enrichment of the plutonium of the MOX fuel will be 
described. 
Since a containing ratio F of a fissionable isotope Puf of plutonium is not 
clear at the design time of the MOX fuel assembly, there is an assumed 
plutonium recovered by reprocessing uranium fuel having a relatively low 
initial enrichment and a small discharge burnup degree, in which the 
ratios of the respective Pu-239, Pu-240 and Pu-241 with respect to the 
entire plutonium are 67%, 22% and 9%, that is, F=76(%), are assumed. 
Although term .DELTA.tc from the uranium fuel discharged time to the 
reprocessing time and term .DELTA.tl from the reprocessing time and the 
manufacturing time are also not clear, these terms are assumed as 
.DELTA.tc=2 (year) and .DELTA.tl=1 (year) from the examples of past 
experience. As described above, these distributions including uranium and 
the plutonium in the present invention are designed to satisfy the thermal 
limit on operational characteristics at the core loading time. 
An embodiment of the present invention at a time when the plutonium is 
actually obtained after the completion of the above design will be 
explained hereunder. 
The actually recovered plutoniums includes plutonium reprocessed by 
reprocessing an initially discharged uranium fuel after the cooling period 
of 4 years therefrom and having the composition of isotope immediately 
after its discharge thereof in which the ratios of Pu-239, Pu-240 and 
Pu-241 with respect to the entire plutoniums are 71%, 20% and 9%, 
respectively, and includes plutonium reprocessed by reprocessing a 
substituted uranium fuel after the cooling period of 4 years therefrom and 
having the composition of isotope immediately after the discharge thereof 
in which the ratios of Pu-239, Pu-240 and Pu-241 with respect to the 
entire plutonium are 47%, 30% and 13%, respectively. The rate in amounts 
thereof is determined to be 1:2. As this result, the average Puf is 
calculated as follows. 
##EQU3## 
In use of such plutonium, when the present invention is applied to the 
correction of the enrichment of the MOX fuel of the present embodiment, 
the following correction will have to be adapted as shown in FIG. 6A. 
##EQU4## 
Namely, in order to manufacture each of the MOX fuel shown in FIGS. 1 to 4 
with the actually obtained plutonium as the fuel maintaining the 
characteristics at the design time, it is necessary to make the enrichment 
Pi of the fissionable plutonium to a value 1.1 times of the enrichment 
concentrations shown in each of the embodiments. 
Further, in the above embodiment, there is described a case where the ratio 
of the obtained plutonium is given, but according to the present 
invention, the mixing ratio for obtaining the plutonium of F=76% as shown 
in the embodiment can be easily obtained. That is, supposing that the 
mixing ratio of the above two kinds of plutonium is Px, the value of the 
Px is as follows. 
##EQU5## 
Next, in the design of the embodiment, .DELTA.tc=2 (year) and .DELTA.tl=1 
(year) are supposed, but with respect to spent fuel due to the actually 
obtained plutonium, it is supposed that .DELTA.tc=4 (year) and the time 
until loading into the core after reprocessing is .DELTA.tl=2 (year). 
Pu-241 at the core loading time at the design time is about 10.0% reduced 
by the amount of 1.7% from 11.7% at the time of .DELTA.tc=0, but in 
practice, the Pu-241 is reduced by the amount expressed by the following 
equation. 
##EQU6## 
Accordingly, since the actually obtained plutonium includes Pu-241 reduced 
by about 3% in amount in comparison with the fissionable plutonium shown 
in the embodiment, the value F is reduced to 64%, and for this reason, it 
is necessary to correct the Pi by 1.05 times. 
Next, in the embodiment, 1 year is assumed for the term .DELTA.tl until 
loading into core after the reprocessing, but a modified embodiment 
corrected by the present invention in consideration of the fact that this 
term becomes 2 years due to the actual MOX fuel manufacturing time. In the 
embodiment, the design is made in consideration of the accumulation of 
Am-241 due to .DELTA.tl =1 (year), but the Am-241 at this time is 
expressed by the following equation, in which Pu-241 is shown as .sup.241 
Pu (ti=0). 
##EQU7## 
In the case after .DELTA.tl=2 (year), Am-241 is increased by, 
##EQU8## 
Now, supposing qm (.DELTA.k/k/Am-241) representing the reactivity effect at 
unit amount of Am-241, the reactivity reduction amount is expressed as, 
##EQU9## 
Then, supposing qn (.DELTA.k/k/Puf) representing the reactivity effect at 
unit amount of Puf, the Puf will be increased by the following amount. 
##EQU10## 
As described hereinbefore, the concentration Pi of the fissionable 
plutonium at the design time is expressed as, 
EQU 1.1.times.1.05.times.fAm.times.Pi (16) 
in consideration of the composition of the actually obtained plutonium, the 
cooling period of spent fuel and the difference between the term until 
loading into core after the reprocessing and the assumed term at the 
design time. 
INDUSTRIAL APPLICABILITY 
As described hereinbefore according to the first embodiment of this 
invention, since the fuel rods containing plutonium are not provided with 
uranium fuel portion and uranium blanket portion, the number of the fuel 
rods manufactured from a predetermined amount of the plutonium can be 
reduced and the effective heat generating length of the fuel rod 
containing the plutonium can be made long, thereby increasing the thermal 
margin to the thermal limit value at the reactor running time due to the 
lowering of the maximum linear power density. 
Further, according to the second embodiment of this invention, since the 
concentration of the plutonium of the fuel rod containing the plutonium is 
set in accordance with the containing rate of plutonium isotope Pu-240, 
the reactivity change caused by the difference of the composition of the 
plutonium isotope can be easily and reasonably compensated for. 
Furthermore, according to the third embodiment of this invention, since a 
predetermined composition of the plutonium isotope is obtained by deciding 
the mixing ratio in accordance with the difference in the Pu-240 amount 
with respect to the plutonium having different isotope composition, the 
concentration can be easily corrected even if the amount of Pu-240 assumed 
at the design time differs from the amount actually obtained. 
Still furthermore, according to the fourth embodiment of this invention, 
since the concentration of the fuel rod containing the plutonium is 
corrected in accordance with the composition of the plutonium, the cooling 
period of spent fuel and the period until the plutonium and the uranium 
are mixed, fabricated and then loaded into the core after the 
reprocessing, more accurate setting of the concentration can be done. 
Still furthermore, according to the fifth embodiment of this invention, 
since the plutonium concentration of the fuel rod containing plutonium is 
set in accordance with a predetermined equation containing the enrichment 
of the uranium fuel, the power peaking factor of the fuel rod containing 
the plutonium can be relatively lowered with respect to that of the 
uranium fuel rod.