Patent Number: 059404612
Section: summary

BACKGROUND OF THE INVENTION The present invention concerns a reactor core for a light water cooled reactor, and a fuel assembly and a control rod constituting the reactor core and, more in particular, it relates to a reactor core for a light water cooled reactor, a fuel assembly and a control rod intended for Pu multi-recycling at a breeding ratio of about 1.0 or slightly greater than 1.0 while keeping the economical and safety performance to the same level as in a BWR (Boiling water reactor) now under operation, that is, while minimizing change of in core structures and maintaining a negative void coefficient. In a nuclear reactor, consumption of fissionable materials such as uranium-235 and plutonium-239 and conversion of fuel fertile materials such as uranium-238 and plutonium-240 into fissionable materials take place by nuclear reactions. A ratio between an amount of fissionable materials contained in fuels taken out of a reactor core and an amount of fissionable materials contained in fuels to be loaded in the reactor core is referred to as a breeding ratio, and the breeding ratio is about at 0.5 in existent light water cooled nuclear reactors. As a method of effectively utilizing uranium resources, it has been considered to increase the breeding ratio. Japanese Patent Laid-Open Sho 55-10591 or Nuclear Technology, vol. 59, 212-227 pp, 1982 disclose that the breeding ratio can be improved by densely arranging fuel rods in a triangular lattice pattern to reduce a water-to-fuel volume ratio in a pressurized water type nuclear reactor. However, the breeding ratio is about 0.9 at the greatest and fissionable materials have to be supplemented for continuous operation without lowering the power. For further increasing the breeding ratio, it may be considered to make a fuel rod gap narrower to further decrease the water-to-fuel volume ratio, but it suffers from a limit, for an embodiment is difficult to attain, in view of manufacture of fuel assemblies and ensurance of thermal margin. On the other hand, Japanese Patent Laid-Open Hei 1-227993 discloses a method of effectively reducing the water-to-fuel volume ratio by utilizing steam voids generated in a reactor core, which is a feature of a boiling water type nuclear reactor. However, it has been shown in the prior art to make the plutonium breeding ratio (a ratio between the amount of fissionable plutonium contained in fuels taken out of a reactor core and an amount of fissionable plutonium contained in fuels loaded in the reactor core; breeding ratio relative to fissionable plutonium) to about 1, but it is not shown to make the breeding ratio (a value is smaller by about 4 to 5% than plutonium breeding ratio in a case of enriching natural uranium with plutonium) to about 1, or 1 or more. If the plutonium breeding ratio is about 1, it is necessary to enrich natural uranium with plutonium for continuing operation without lowering the power and the uranium resources can not be used up thoroughly. In the present invention, the breeding ratio of about 1 means a value 0.98 or more. SUMMARY OF THE INVENTION A first object of the present invention is to provide a reactor core and a fuel assembly capable of maintaining the power generation cost, thermal margin and safety about to the same level as in light water cooled reactors now under operation, thereby contributing to stable long time energy supply. A second object of the present invention is to provide a reactor core, a fuel assembly and a control rod for attaining a breeding ratio at 1.0 with a Pu-enriched degraded uranium fuel by reducing the water-to-fuel volume ratio for contributing to stable long time energy supply. A third object of the present invention is to provide a reactor core and a fuel assembly capable of operating as many nuclear power reactors as possible with a predetermined amount of Pu by reducing a required Pu inventory per unit power for contributing to stable long time energy supply. A fourth object of the present invention is to provide a reactor core and a fuel assembly capable of attaining the same extent of power and same extent of burnup degree with the same extent of thermal margin by using the same materials and about the same size of a pressure vessel as those in reactors now under operation for making the power generation cost about equal with that in the existent light water cooled reactors. A fifth object of the present invention is to provide a reactor core and a fuel assembly capable of attaining a negative void coefficient by increase of neutron leakage in the direction of a reactor core height and power distribution swing in the direction of the reactor core height upon power up in order to make the safety to the same extent as in existent light water cooled reactors. A sixth object of the present invention is to provide a reactor core capable of confining radioactivated materials present in the reactor within a pressure vessel by maintaining a distillation function due to boiling in order to make the safety to the same extent as the existent light water cooled reactors. A seventh object of the present invention is to provide a reactor core and a fuel assembly capable of recycling Pu and U simultaneously while abolishing sole Pu extraction for coping with nuclear nonproliferation. An eighth object of the present invention is to provide a reactor core and a fuel assembly for causing actinoid nuclides together with uranium and plutonium in the reactor and recycling them in order not to leave long life radioactive wastes in future generations. For attaining the first object, there is provided in accordance with the present invention, a reactor core having fuels comprising uranium containing at least one of degraded uranium, natural uranium, depleted uranium and low concentrated uranium, enriched with Pu or Pu and actinoid nuclides, wherein a breeding ratio is about 1.0, or 1.0 or more and a void coefficient is negative. Further, for attaining the first object, there is provided in accordance with the present invention, a fuel assembly having fuels comprising uranium containing at least one of degraded uranium, natural uranium, depleted uranium and low concentrated uranium, enriched with Pu or Pu and actinoid nuclides, wherein a breeding ratio is about 1.0, or 1.0 or more. Further, for attaining the second object, there is provided in accordance with the present invention, a hexagonal dense fuel assembly comprising fuel rods arranged in a regular triangular lattice pattern, wherein a gap between the fuel rods is from 0.7 to 2.0 mm, as well as a reactor core constituted with the fuel assembly. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core and a fuel assembly wherein an effective water-to-fuel volume ratio is between 0.1 and 0.6. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor wherein an average envichment of fissionable Pu in a reactor core is from 6 to 20% except for outer circumferential portion and blanket portions in upper and lower ends of a reactor core. Further, for attaining the second object, there is provided in accordance with the present invention, a fuel assembly wherein an average enrichment of fissionable Pu is from 6 to 20% in a region excepting for blanket portions at both of upper and lower ends. Further, for attaining the second object, there is provided in accordance with the present invention, a boiling water type light water cooled reactor wherein an average void fraction in the reactor core is from 45 to 70% during operation at 50% or more rated power. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor comprising hexagonal fuel assemblies and a cluster-type control rod inserted therein. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor comprising hexagonal fuel assemblies and Y-type control rods each inserted therein and having three wings at 120 degree spacing between each of the wings. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor, wherein the Y-type control rods adjacent to a regular hexagonal fuel assembly each has two or less wings, and a gap between fuel assemblies not inserted with the wing between the fuel assemblies is narrower than a gap between the fuel assembles inserted with the wing. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor comprising square fuel assemblies arranged densely in a regular triangular lattice pattern and cross type control rods each having four wings to be inserted therebetween at a spacing of 90 degree between each other wings. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor having eclipsed hexagonal fuel assemblies in which fuel rods are arranged in a regular triangular lattice pattern, wherein the number of rows of two sets of fuel rods is equal to each other and is greater by one row than the number of the rows of the remaining one set of fuel rods among three sets of rows of fuel rods in parallel with opposing rows of fuel rods at the outermost layer, and a regular hexagonal fuel assembly lattice is constituted of the eclipsed hexagonal fuel assembly and a wing of the Y-type control rod. Further, for attaining the second object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor comprising a cluster type, Y-type or cruciform control rod having a follower portion at the top end of the control rod made of a material having a smaller moderating function than that of light water, for embodiment, carbon, heavy water, beryllium, Zr alloy or stainless steel for excluding moderators. For attaining the second object, there is provided in accordance with the present invention, a hexagonal fuel assembly and an eclipsed hexagonal fuel assembly wherein at least two or more regions ranging from a region adjacent to the Y-type control rod to a region apart from the Y-type control rod are constituted of plural kinds, particularly, two to five kinds of fuel rods of different enrichments of fissionable Pu. For attaining the second object, there is provided in accordance with the present invention, a square fuel assembly, wherein at least two or more regions ranging from a region adjacent to the cruciform control rod to a region apart from the cruciform control rod are constituted with plural kinds, particularly, two to five kinds of fuel rods of different enrichments of fissionable Pu. For attaining the third, fourth and fifth objects, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor wherein an average power density in the reactor core is from 100 kW/l to 300 kW/l excepting for blanket portions at the outer circumference and at upper and lower ends of a reactor core. For attaining the third, fourth and fifth objects, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor wherein a portion having an average enrichment of fissionable Pu along the horizontal cross section of the fuel assembly of 6 w/o or more is between 40 cm and 140 cm with respect to the axial direction excepting for the blanket portions at both of upper and lower ends of the reactor core. For attaining the third, fourth and fifth objects, there is provided in accordance with the present invention, a fuel assembly wherein a portion having an average enrichment of fissionable Pu along the horizontal cross section of the fuel assembly of 6 w/o or more is between 40 cm and 140 cm with respect to the axial direction of the fuel assembly excepting for the blanket portions at both of upper and lower ends. Further, for attaining the fourth object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor wherein the reactor core is bisected radially into equal areas except for the outermost circumference of the reactor core and fuel assemblies are loaded such that an average value for the number of core staying cycles of the fuel assemblies in the outer reactor core region is made smaller than that in the inner reactor core region. Further, for attaining the fourth object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor wherein an average orifice pressure loss coefficient of fuel assemblies at or adjacent to the outermost circumference of a reactor core is greater than an average orifice pressure loss coefficient in other regions. Further, for attaining the fifth object, there is provided in accordance with the present invention, a hexagonal fuel assembly wherein an average value of the enrichment of fissionable Pu in a lower half portion is less than an average value in an the upper half portion excepting for the blanket portions at both of upper and lower ends. Further, for attaining the fifth object, there is provided in accordance with the present invention, a fuel assembly wherein portions having an enrichment of fissionable Pu of 6 w/o or more are disposed in upper and lower portions along the axial direction of the fuel assembly excepting for the blanket portions at both of the upper and lower ends, and the enrichment of fissionable Pu in a region therebetween near a central portion is 6 w/o or less. Further, for attaining the sixth object, there is provided in accordance with the present invention, a reactor core for a boiling water type light water cooled reactor, wherein the steam weight ratio of coolants at an exit of the reactor core is from 20% to 40% during operation at 50% or more rated power. Further, for attaining the seventh object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor and a fuel assembly, wherein Pu and uranium are recycled simultaneously. Further, for attaining the eighth object, there is provided in accordance with the present invention, a reactor core for a light water cooled reactor and a fuel assembly, wherein Pu, uranium and actinoids are recycled simultaneously. According to the studies made by the present inventors of the application, the following have been found. The amount of natural uranium resources in the world is estimated to be about 15,000,000 tons, which corresponds to an amount capable of operating 1,000 units of existent light water cooled reactors with 1,000,000 kW of electric power for about 100 years. As a result, degraded uranium of nearly about 15,000,000 tons and fissionable Pu of 15,000 tons will be left. Accordingly, a power reactor (RBWR) at a breeding ratio of 1.0 can continue nuclear fission for degraded uranium using fuels containing Pu-enriched degraded uranium under the catalyst-like effect of Pu with inside and outside inventory of fissionable Pu at 10 tons per 1,000,000 kW electric power. Since uranium generates a heat energy of about 1 MWD per 1 g, 1,500 units of RBWR can be operated for 10,000 years and the entire uranium resources can be used up thoroughly so that it can contribute to the stable long time supply for the first object. Further, the second object can be attained by the functions described below. According to the studies of the present inventors, for the relationship between the breeding ratio and the effective water-to-fuel volume ratio in the reactor core of the light water cooled reactor, the followings have been found. The effective water-to-fuel volume ratio (Vm/Vf)eff! is extended from the geometrical water-to-fuel volume ratio (Vm/Vf)geo; water-to-fuel volume ratio not generating steam void! considering generation of steam void in the reactor core. Assuming the decreasing ratio of hydrogen density due to generation of the steam void as F, there is the following relationship between both of them. EQU (Vm/Vf)eff=F(Vm/Vf)geo (equation 1) Further, F has the following relationship with the average steam void fraction V (%) for reactor core. EQU F=(100-V)/100+f V/100 where F is the ratio of saturated steam density to saturated water density. PA1 .alpha.: number of new neutrons generated when a neutron is absorbed in a fissionable material and one fissionable material is annihilated. PA1 .beta.: additional contribution by nuclear fission of fuel fertile material in a high speed energy region. PA1 .gamma.: ratio of wasteful neutron capture relative to neutron absorption amount by fissionable material (including neutron leakage). Generally, f is a value as small as about 1/20 and F can be approximated as below. EQU F=(100-V)/100 FIG. 2 shows a conversion ratio defined from an effective water-to-fuel volume ratio and neutron balance, and relationship between each of three factors constituting the conversion ratio. EQU Conversion ratio=.alpha.(1+.beta.)-(1+.gamma.) (equation 2) where In the light water cooled reactors now under operation, the effective water-to-fuel volume ratio is about 2.0 and the breeding ratio is about 0.5. For attaining the breeding ratio at about 1, it is necessary to make the conversion ratio to about 1. According to the studies of the present inventors, it has been found that the breeding ratio of about 1 at a conversion ratio of 0.85 or more can be achieved by increasing the enrichment of fissionable Pu within a range to be described later and increasing the neutron leakage to the blanket. The effective water-to-fuel volume ratio for this purpose is 0.6 or less. On the other hand, for obtaining the effective fuel-to-water volume ratio of 0.1 or less, the average steam void fraction for the reactor core has to exceed 70%, so that a two phase flow state can no more be maintained at the exit of the reactor core upon transient event. The effective water-to-fuel volume ratio of 0.1 to 0.6 can be attained by densely arranging fuel rods, utilizing the steam void generated in the reactor core or, if the control rod is not inserted, by inserting a follower at a control rod insertion position to exclude the moderator, or by the combination of such three means. FIG. 3 shows an embodiment of a relationship between the fuel rod gap and the geometrical water-to-fuel volume ratio. In FIG. 3, the diameter of the fuel rod is within a range about from 9.5 to 12.3 mm which is used at present in light water cooled reactors and a regular triangular fuel rod lattice is adopted. If the fuel rod gap is reduced to 2 mm or less, (Vm/Vf)geo of the fuel rod lattice is about 0.9 or less. In a case of a fuel assembly having fuel rods arranged densely in a regular triangular lattice pattern, (Vm/Vf)geo of the fuel assembly takes a value greater by 0.1 to 0.2 than (Vm/Vf)geo of the fuel rod lattice, in view of a gap region or a control rod insertion region between the fuel assemblies. Accordingly, for attaining the effective water-to-fuel volume ratio of 0.6 or less under the geometrical water-to-fuel volume ratio, it is necessary to make the average steam void fraction for the reactor core to 45% or more in view of equation 1 (steam weight ratio at the exit of the reactor core is 20% or more in view of the relationship shown in FIG. 27). On the other hand, at a fuel rod gap within a range from 0.7 (minimum value for the fuel rod gap considering manufacture of the fuel assembly and ensurance for the thermal margin) to 1.0 mm (it can be more than 1.0 mm in a case where the diameter of the fuel rod is greater than 9.5 mm), (Vm/Vf)geo can be reduced to about 0.6 or less at the steam void fraction of 0%. FIG. 4 shows a relationship between the average enrichment of fissionable Pu of the fuel assembly and the breeding ratio. For maintaining the reactor core in a critical state throughout the operation period, it is necessary to increase the fissionable Pu enrichment to 6 wt % or more. On the other hand, while the breeding ratio decreases along with the fissionable Pu enrichment it has been found that the breeding ratio of about 1 can be attained up to 20 wt % by utilizing the increase of the excess reactivity and increasing the neutron leakage to the blanket as described above. Further, in this case, as the means for controlling the reactivity, a method may be considered of inserting a cluster-type control rod into the fuel assembly, or inserting a Y-type control rod at the periphery of the hexagonal fuel assembly or inserting a cruciform control rod at the periphery of the square fuel assembly. A reactor core having a breeding ratio of 1.0 can be attained by the combination of the foregoing means. Further, the third, fourth and fifth objects can be attained by the following functions. According to the studies of the present inventors, it has been found that the height of the fuel assembly (effective reactor core length: length of a region having an average fissionable Pu enrichment of 6 wt % or more along a horizontal cross section) can be reduced while ensuring thermal margin by setting the power of fuel assemblies per unit horizontal cross section of the reactor core to the same extent as that in the existent boiling water type light water cooled reactor. As a result of densely arranging the fuel rods for reducing the effective water-to-fuel volume ratio to 0.6 or less, the number of fuel rods per unit horizontal cross section of the reactor core is 3 to 4 times of the existent boiling water type light water cooled reactor. Accordingly, the height of the fuel assembly (effective reactor core length) providing the same extent of average linear power density is about 1/3 to 1/4 of the existent boiling water type light water cooled reactor. Further, since the moderators are homogeneously dispersed as compared with the existent boiling water type light water cooled reactor, the local power peaking coefficient of the fuel rod can be reduced by about 30% or more (by using the enrichment distribution if necessary). Further, the power peaking coefficient can be reduced by about 40% or more due to less change of burning reactivity and void reactivity and further in combination with other means to be described later. Accordingly, the height of the fuel assembly (effective reactor core length) providing an average linear power density equal to or greater than that of the existent boiling water type light water cooled reactor is 40 cm or more, which is about 1/10 thereof. On the other hand, by making the effective reactor core length shorter to increase the axial neutron leakage, the effect of reducing the void coefficient can be utilized. It has been found according to the studies of the present inventors, that negative void coefficient can be achieved in combination with other means described later if the effective reactor core length is set to 140 cm or less. While the power generation ratio in the blanket portion is increased by reducing the length, the average power density in the region excepting for the blanket portion is about 100 to 300 kW/l by the reduction of the effective reactor core length. As a result, since RBWR producing the same power can be contained within a pressure vessel having the diameter about equal to that in the existent reactor, the power generation cost can be kept about the same as that in the existent light water cooled reactor and the safety can also be kept at a level about equal to that in the existent light water cooled reactor. Further, this can decrease the Pu inventory and, accordingly, a number of power reactors can be operated by a predetermined amount of Pu to attain stable energy supply. Further, the fifth object can be attained by the following functions. According to the studies of the present inventors, the axial power distribution of the reactor core can be flattened by increasing the fissionable Pu enrichment in the upper portion of the reactor core to greater than that in the lower portion of the reactor core and, as a result, the Pu inventory can be decreased. Further, while the steam void fraction in the reactor core is increased upon power up or lowering of the reactor core coolant flow rate as shown in FIG. 5 ,the power distribution swings to the lower portion of the reactor core where the fissionable Pu enrichment is relatively low and neutron importance is small thereby enabling to reduce the reactivity of the reactor core (negative void coefficient). Further, the second object can be attained by the following functions. According to the studies of the present inventors, a reactor core constitution in which fuel rods are arranged densely in a regular triangular lattice pattern can be attained by inserting a cluster-type control rod in the fuel assembly. Further, according to the studies of the present inventors, a reactor core constitution in which fuel rods are arranged densely in a regular triangular lattice pattern can be attained also by combining the hexagonal fuel assembly and a Y-type control rod. The Y-type control rod and the hexagonal fuel assembly can be combined by a method of making a fuel assembly into a regular hexagonal shape and by a method of constituting a regular hexagonal shape with one wing of the Y-type control rod and a fuel assembly. The former has a merit capable of simplifying the constitution of the fuel assembly and the latter has a merit of making the central position of the assembly in the reactor core as a regular triangular shape. Further, according to the studies of the present inventors, a reactor core constitution in which fuel rods are arranged densely in a regular triangular lattice pattern can be attained also by combining a square fuel assembly with a cruciform control rod. Further, according to the studies of the present inventors, in a combination of the hexagonal fuel assembly and the Y-type control rod or a combination of the square fuel assembly and the cruciform control rod, a neutron moderating effect of water is increased upon withdrawal of the control rod for the fuel rod facing the control rod, so that the neutron energy is reduced and, if the fissionable Pu enrichment for each of the fuel rods in the assembly is made identical, a power peaking is generated to the fuel rods facing the control rod. Then, a fuel assembly in which the power distribution in the fuel assembly is flattened can be attained by varying the fissionable Pu enrichment in the assembly for several kinds in accordance with the distance from the insertion position of the control rod. Further, the fourth object can be attained by the following functions. According to the studies of the present inventors, the power and flow rate of the fuel assembly can be flattened to improve the thermal margin by making the arrangement of the fuel assemblies and the orifice constitution appropriate in the reactor core. The infinite neutron multiplication factor in the outer reactor core region can be made higher than that in the inner region to flatten the radial power distribution by radially bisecting the region of the reactor core except for the outermost circumference into equal areas by loading the fuel assemblies such that the average value for the number of core staying cycles of the fuel assemblies in the outer reactor core region is made smaller than that in the inner reactor core region. Necessary reduction for the fissionable Pu enrichment can be attained by loading fuel assemblies of greater number of staying cycles in the outermost circumferential region of the reactor core at a low neutron importance. According to the studies of the present inventors, the effect of the neutron leakage from the outer circumferential portion of the reactor core is particularly remarkable in the fuel assembly at and adjacent to the outermost layer of the reactor core, to lower the power of the fuel assembly compared with that in other regions and increase the flow rate flowing in the fuel assembly. Accordingly, the distribution of the flow rate can be flattened by setting such that the average value for the orifice pressure loss coefficient of fuel assemblies at and adjacent to the outermost circumference of the reactor core is made greater than the average value for the orifice pressure loss coefficient in other regions. This can reduce the flow rate near the outermost circumference of the reactor core and reduce the flow rate over the entire reactor core. Further, the steam void fraction can be increased in the region where the orifice pressure loss coefficient is increased, thereby contributing to the improvement of the void coefficient and increase of the breeding ratio. Further, the sixth object can be attained by the following functions. According to the studies of the present inventors, while the breeding ratio can be increased to 1.0 or more by cooling light water steams, development of a material having greater high temperature resistance than that used in the existent BWR is necessary since the steam temperature exceeds a saturation temperature, and radioactive nuclides such as corrosion products flow together with the steams out of the reactor core. In the present invention, since the steam weight ratio at the exit of the rector core is kept to 40% or less, by which the coolants keep a two phase flow state at the saturation temperature even upon power up caused by abnormal transient change, to maintain the saturation temperature, thereby enabling use of the same structural materials as those in the existent light water cooled reactor, and a reactor core at a breeding ratio of 1.0 or more can be attained while preventing incorporation of radioactive nuclides such as corrosion products in the steams sent to the turbine by the distillation function due to boiling in the rector core. Further, the first and the seventh objects can be attained by the following functions. The present inventors, have studied embodiments regarding fuels comprising degraded uranium enriched with Pu generated as residues upon manufacture of enriched uranium used in the existent light water cooled reactors, intended for stable long time energy supply. If natural uranium or depleted uranium recovered from spent fuels are still present in a great amount as at the present situation, a reactor core having equal or higher performance with respect to the breeding ratio and the void coefficient as compared with the case of Pu-enriched degraded uranium can be attained by reducing the enrichment of the fissionable Pu by about 0.5 wt % or more as compared with a case of using degraded uranium, by enriching Pu to natural uranium, depleted uranium or low concentrated uranium (0.71 wt % -2.0 wt %) instead of degraded uranium. Further, the eighth object can be attained by the following functions. According to the studies of the present inventors, long life radioactive nuclides are in an equilibrated state in the reactor, to reach a predetermined amount not only by enriching degraded uranium with Pu but also by recycling actinoid nuclides simultaneously. Accordingly, in the reactor according to the present invention, the amount of generation and the amount of annihilation are equilibrated for the actinoid nuclides to reduce the increment to zero thereby enabling to attain a nuclear reactor system capable of not only remarkably reducing the entire generation amount of the long half-life actinoid nuclides that particularly result in problems, among the radioactive wastes, and but also confining Pu-containing actinoid nuclides only within the nuclear reactor, reprocessing facility and fuel manufacturing facility. Followings are effects of the present invention in accordance with above-mentioned features of the present invention. According to the present invention, degraded uranium, natural uranium, depleted uranium or low concentrated uranium can be burnt under a catalyst-like effect of Pu by attaining a breeding ratio of about 1.0, or 1.0 or more using fuels formed by adding Pu to degraded uranium, natural uranium, depleted uranium or low concentrated uranium, thereby enabling to contribute to stable long time energy supply. Further, since the effective water-to-fuel volume ratio of 0.1 to 0.6 is provided by the combination of the dense hexagonal fuel assembly or the square fuel assembly comprising fuels formed by adding Pu to degraded uranium, natural uranium, depleted uranium or low concentrated uranium, coolants at high void fraction of 45% to 70% and the cluster-type, Y-type or cruciform control rods, a breeding ratio of about 1.0, or 1.0 or more can be attained, thereby enabling to contribute to the stable long time energy supply. Further, the Pu inventory can be decreased by obtaining the same power as that in ABWR now under construction using the pressure vessel of substantially identical diameter and a short fuel assembly providing the reactor core height of 40 to 140 cm, so that a number of reactors according to the present invention can be operated with Pu generated from spent fuels from light water cooled reactors, under restricted natural uranium deposits in the world, thereby enabling to contribute to the stable long time energy supply. Further, since the diameter of the pressure vessel, the operation conditions such as power and the materials used are substantially identical with those in BWR now under operation, the power generation cost can be kept about to an identical level as that in existent BWR although the performance is greatly improved. Further, since increase of the neutron leakage in the vertical direction of the reactor core and the swing of the power distribution in the vertical direction of the reactor core can be utilized effectively by using the short fuel assembly, the upper and lower two region fuel assembly and the axially inhomogeneous fuel assembly, a reactor core of negative void coefficient can be attained to provide the same extent of safety as in existent fuel once-through type light water cooled reactors. Further, since the combination of the dense fuel assembly and the coolants at high void fraction can increase the ratio of neutrons in the resonance energy region, increase the Doppler effect and decrease the absolute value of the negative void coefficient, safety can be improved such as for the events of power up, pressurization, decrease of coolant void fraction. Further, since the steam weight ratio of the coolants at the exit of the reactor core is kept to 40% or less, radioactive materials such as corrosion products accumulated in the reactor can be confined within the reactor by maintaining the distillation function due to boiling, thereby enabling to maintain the radioactive level on the side of the turbine to the same extent as that in BWR now under operation and remarkably reduce the radiation level over that in existent vapor cooled fast reactors in the existent concept for the breeding reactor. Further, the reactor core comprising the hexagonal fuel assembly and the cluster type control rod inserted therein can increase the reactor core homogeneity and enhance the thermal margin. Further, the reactor core comprising the hexagonal fuel assembly and the Y-type control rod inserted between the fuel assemblies can utilize the technique in existent BWR of inserting from the lower portion of the reactor core as it is. Further, the reactor core comprising the square fuel assembly and the cruciform control rod inserted between the fuel assemblies can utilize the reactor core system in existent BWR as it is. Further, in the hexagonal or eclipsed hexagonal fuel assembly or the square fuel assembly, constitution of multiple regions, particularly, 2 to 5 regions ranging from a region in the vicinity of the Y-type or cruciform control rod to a region apart from the control rod with 2 to 5 kinds of fuel rods having varied fissionable Pu enrichments can reduce the power peaking in the fuel assembly and enhance the thermal margin. Further, increase of the reactor core power density from 100 to 300 kW/l can reduce the Pu inventory amount per unit power and increase the capacity of the power generation facilities of the present invention that can be operated for a predetermined Pu, thereby contributing to the stable long time energy supply. Further, since the portion having the average enrichment of the fissionable Pu along the horizontal cross section of 6 w/o or more is between 40 to 140 cm in the axial direction of the fuel assembly, the Pu inventory amount per unit power is decreased, the capacity of the power generation facilities of the present invention that can be operated for predetermined Pu can be increased thereby contributing to the stable long time energy supply, as well as the neutron leakage effect in the direction of the reactor core height is increased when the amount of steams generating is increased to make the negative void coefficient greater and contribute to the safety. Further, since the average value for the fissionable Pu enrichment in the lower half portion of the fuel assembly using blanket portions at both of upper and lower ends is lower than that in the upper half portion, the power distribution in the direction of the reactor core height is flattened to enhance the thermal margin, as well as the power distribution swings in the direction of the reactor core height upon increase of the steam generation amount, which makes the negative void reactivity coefficient greater to contribute to the safety. Further, since portions having the fissionable Pu enrichment of 6 w/o or more are provided in the upper and the lower portions along the axial direction excluding the blanket portions at both of the upper and lower ends of the fuel assembly, and the fissionable Pu enrichment in a central region therebetween is reduced to 6% or less, the reactor power is increased, the negative void coefficient due to the swing of the power distribution in the vertical direction of the reactor core is increased when the steam amount in the reactor core is increased to improve the safety. Further, the neutron absorbing effect in the region near the axial center of the reactor core can increase the Pu inventory capable of being loaded to the reactor core to improve the function as the Pu storing reactor. Then, since the reactor core height is relatively increased and the entire length of the fuel rod is increased, the thermal margin relative to the maximum linear power density can also be improved. Further, simultaneous recycling of Pu and uranium can increase the preventive effect regarding nuclear nondiffusion. Further, simultaneous recycling of Pu, uranium and actinoid nuclides can provide balance between the amount of generation and annihilation of the actinoid nuclides to reduce the increment to zero, as well as long half-life actinoid nuclides that result in problems, particularly, among radioactive wastes, can be confined only within the reactor, reprocessing facility and fuel production facility to improve performance for environments.