Patent Number: 
Section: description

The boiling water reactor of a first embodiment according to this invention is hereafter explained with reference to FIG. 1. FIG. 1 is a graph showing a relation between xcex94MCPR and a dynamic void coefficient, during a pressure rise transient and a water-supply transient, which are abnormal transient changes during operating power in a natural-circulation water reactor. During operating power, an average void fraction of 40% is assumed here. Full-capacity of a by-pass is assumed for the pressure rise transient. That is, when the full-capacity of the by-pass is employed and a turbine governor valve carries out rapid closing, rapid opening of a turbine by-pass valve is achieved. About 100% of main steam flow is bypassed and processed to a condenser. The full capacity which is -bypassed is equal to about 100% of the main steam flow. The solid line with attached symbol 110a of the figure shows xcex94MCPR during the pressure rise transient. Moreover, the dashed line with attached symbol 110b shows xcex94MCPR during a water-supply temperature reduction transition (due to, for example, improper operation of water-supply temperature control equipment). Point P in the FIGURE shows the point that xcex94MCPR of both transients become the same value. The dynamic void coefficient at this time is about xe2x80x9cxe2x88x925¢xe2x80x9d. Here, xe2x80x9c¢xe2x80x9d is a unit of reactivity. As shown in FIG. 1, in the reactor core where the dynamic void coefficient is small (the absolute value of the dynamic void coefficient is large), xcex94MCPR of the pressure rise transient exceeds xcex94MCPR of the water-supply transient. That is, when the dynamic void coefficient is increased (i.e. an absolute value is decreased), xcex94MCPR for both transients can be decreased. In this embodiment, the xcex94MCPR of the pressure transient is made completely below xcex94MCPR of the water-supply transient by ensuring that the dynamic void coefficient under the above situations is larger than xe2x80x9cxe2x88x925¢xe2x80x9d. That is, an absolute value of the dynamic void coefficient is less than xe2x80x9c5¢xe2x80x9d (equivalent to a delayed neutron fraction being less than 5%). Therefore, the xcex94MCPR and the reactor core stability are improved by making the dynamic void coefficient of the reactor core larger (i.e. the absolute value of the dynamic void coefficient small) than xe2x80x9cxe2x88x925¢xe2x80x9d. The dynamic void coefficient is defined according to the following equation.       dynamic    ⁢          xe2x80x83        ⁢    void    ⁢          xe2x80x83        ⁢    coefficient    =            (                        void          ⁢                      xe2x80x83                    ⁢          reactivity          ⁢                      xe2x80x83                    ⁢          coefficient                          delayed          ⁢                      xe2x80x83                    ⁢          neutron          ⁢                      xe2x80x83                    ⁢          fraction                    )        ·          (              average        ⁢                  xe2x80x83                ⁢        void        ⁢                  xe2x80x83                ⁢        fraction            )       The average void fraction is usually 0.4. It is known that the delayed neutron fraction xcex2 is about 0.006 in a nuclear reactor using a uranium fuel. Thus, a dynamic void coefficient of xe2x88x925¢ is equivalent to about xe2x88x920.07 (ratio of % xcex94k/k/% void) of the void reactivity coefficient. Although the value of the void reactivity coefficient changes depending on the state of the reactor core, the void reactivity coefficient in the reactor core of first loading is xe2x88x920.03% xcex94k/k/% void, which is about half of the void reactivity coefficient under normal operation. Therefore, setting the dynamic void coefficient larger than xe2x88x925¢ is the same as the void reactivity coefficient satisfying the following formula: xe2x88x920.07% xcex94k/k/% void ratioxe2x89xa6the void reactivity coefficientxe2x89xa6xe2x88x920.03% xcex94k/k/% void ratio. Here, the delayed neutron and delayed neutron fraction are explained. It takes time until a neutron is generated after nuclear fission happens. A neutron generated some time after a fission reaction is called a delayed neutron. The delayed neutron fraction xcex2 represents the ratio of the number of delayed neutrons and the whole number of generating neutrons which are generated in nuclear fission. In the first embodiment, the width of the by-pass portion is increased and the absolute value of the dynamic void coefficient is reduced to ensure that the void reactivity coefficient is set within the above mentioned range, to promote a slowdown of the neutrons. That is, the absolute value of the dynamic void coefficient becomes small by softening the neutron spectrum. If the value of the width of the by-pass portion divided by the width of the channel box is 0.12 or more, since a sufficient neutron slowdown effect will be obtained, the dynamic void coefficient can be made larger (i.e. the absolute value of the dynamic void coefficient small) than or equal to xe2x88x925¢. The width of the by-pass portion will be further explained referring to FIG. 2. In this embodiment, since the dynamic void coefficient is adjusted by the width of the by-pass portion, the core of the fuel assembly does not need to be changed. Furthermore, the structure of the fuel assembly does not need to be changed. FIG. 2 is a graph showing the correlation between the ratio (namely, value of d2/d1) of the width d2 of the by-pass portion 7 to the width d1 of the channel box 4 of a fuel assembly (similar to the one shown in FIG. 13), and dynamic void coefficient. According to FIG. 2, when the ratio is 12% or more, it turns out that the dynamic void coefficient is xe2x88x925¢ or more. Therefore, proper selection of the width d2 of the by-pass portion 7 with respect to the width d1 of the channel box 4 of the fuel assembly, in this embodiment, make the dynamic void coefficient larger than xe2x88x925¢. This means that d2/d1 greater than =0.12 is required. Since d2/d1 less than =0.10 in the conventional nuclear reactor, when the same fuel assembly as the conventional nuclear reactor is used, in this embodiment the width of the by-pass portion is designed to be 20% or more larger than the conventional nuclear reactor. By expanding the by-pass area, the reactivity at cold temperatures is reduced, and reactor shutdown margin is improved. In this first embodiment, by setting up the width of the by-pass portion greater than the conventional nuclear reactor, the void coefficient can be adjusted in the suitable range. Therefore, the core of the fuel assembly and the channel box width do not need to be changed, and the performance of the natural circulation reactor is improved even though a fuel assembly which has the same structure as the conventional nuclear reactor is used. Furthermore, in this embodiment, by expanding the by-pass area, the reactivity at cold temperatures is reduced, and shut down margin is improved. The boiling water reactor of the second embodiment according to this invention is explained hereafter. In this embodiment, like the first embodiment, the dynamic void coefficient during operating power is greater than xe2x88x925¢ with a 40% average void fraction. As a result, pressure transient characteristics and in-core stability are improved. In the first embodiment, the width of the by-pass portion between fuel assemblies is adjusted. In this embodiment, to replace the method of the first embodiment and to provide a dynamic void coefficient greater than xe2x88x925¢, the axial enrichment distribution of the fuel assembly is adjusted. FIG. 3 shows the axial enrichment distribution of a fuel effective section of a fuel assembly in this embodiment. As shown in FIG. 3, the fuel effective section of the fuel assembly is divided into two vertical areas in which the enrichment of uranium 235 differs. A symbol 31 shows an upper area and a symbol 32 shows a lower area. The axial length of the upper area 31 and the lower area 32 are expressed as dU, dL in FIG. 3, respectively. Moreover, the enrichment of uranium 235 is expressed as eU, eL, respectively. In this embodiment, the enrichment of the upper area is 0.3 wt % higher than the enrichment of the lower area. Thereby, the peak position of axial power is contained in the upper area. Furthermore, in consideration of the peak position, the axial length of the lower area 32 is within the limits of one-third to one-half of the fuel effective section full length. That is, ⅓ less than =dL/(dL+dU) less than =xc2xd. Generally, a boiling water nuclear reactor has the property that the void fraction increases toward the upper part at the time of output operating power. Therefore, a neutron slowdown is achieved in the lower part and power also tends to become large. Therefore, power in the upper part can be gradually increased by increasing the upper enrichment from the lower part. Moreover, if the difference between the enrichment in the upper part and the enrichment in the lower part is 0.3 wt % or more, the power of the reactor core can always maintain a peak in the upper part during operating power. The average void fraction of a reactor core with a power peak always in the upper part becomes smaller than the average void fraction of a reactor core which has a power peak in the lower part. Because the absolute value of the dynamic void coefficient becomes small in accordance with the reduction of the average void fraction, substantially the same effect as making the absolute value of the dynamic void coefficient small as in the first embodiment can be obtained. In a nuclear reactor core loaded with fuel which has the above-mentioned structure, the peak position of the axial power is in the upper area of the fuel effective section. Therefore, the same effect as having made the average void fraction of the reactor core low, and having made the absolute value of the dynamic void coefficient substantially small, is obtained. Consequently, in-core stability during a pressure rise transient are further improved. Moreover, in this embodiment, by decreasing the average void fraction, the pressure loss of the reactor core is reduced and the natural circulating flow rate increases in accordance with decreasing the average void fraction. Furthermore, xcex94MCPR improves. As noted above, in the fuel assembly loaded in the conventional nuclear reactor, the difference between the enrichment in the upper part and the enrichment in the lower part is about 0.2 wt % or less. When the enrichment of the reactor core upper area is increased such that the enrichment difference exceeds 0.2 wt %, when a xe2x80x9cshutdownxe2x80x9d occurs, there is not a comfortable margin of safety. On the other hand, in this embodiment, the absolute value of the dynamic void coefficient is made 5¢ or less. Since the reactor shut down margin has been sharply improved by setting the absolute value to be 5¢ or less, this allows making the enrichment difference more than 0.3 wt %. Although FIG. 3 shows two different types of enrichment, the invention is not limited to two different types. Even if three or more kinds of enrichments are used, the same action and same effect as above are obtained by making the difference between a maximum enrichment of an upper area, and a minimum enrichment of a lower area more than 0.3 wt %. The boiling water reactor of a third embodiment according to this invention is explained hereafter. In this embodiment, a dynamic void coefficient is obtained similar to the first two embodiments, greater than xe2x88x925¢ with a 40% average void fraction during power operation. In the second embodiment, the enrichment distribution in the vertical direction of the fuel effective section of the fuel assembly is adjusted. In this embodiment, the enrichment distribution at the vertical ends in the axial direction of the fuel assembly is designed to replace the second embodiment and to provide a dynamic void coefficient greater than xe2x88x925¢. FIG. 4 shows the distribution of blanket areas of the vertical ends in the axial direction. As shown in FIG. 4, an upper blanket area 33 is located at a top edge of the fuel assembly and a lower blanket area 34 is located at a lower edge. These areas have a low enrichment eN which consists of natural uranium or depleted uranium. The axial length of the upper blanket area 33 and the lower blanket area 34 are expressed as fU, fL, respectively. The area inserted between these blanket areas 33 and 34 is the ordinary enriched-uranium area 35. In this embodiment, the length a of the lower blanket area 34 is greater than the length fU of the upper blanket area 33. In the conventional boiling water nuclear reactor, the length of the lower blanket is less than the length of the upper blanket. By making the lower blanket area 34 relatively longer, the peak position of the axial power output will be in the upper part of the fuel assembly. Therefore, natural circulation flow will increase, the average void fraction becomes low, and the absolute value of the dynamic void coefficient is reduced. Therefore, xcex94MCPR of the pressure rise transient and a delay ratio of in-core stability are further improved. Furthermore, in this embodiment, since the average void fraction decreases, the pressure loss of the reactor core declines. Since natural-circulation flow rate increases in connection with the reduction of the reactor core pressure loss, xcex94MCPR improves. The boiling water reactor of a fourth embodiment according to this invention is explained hereafter. In this embodiment, the dynamic void coefficient during rated (normal) power operation is the same as the first three embodiments, that is, greater than xe2x88x925¢ and an average void fraction of 40%. In the fourth embodiment, a water-rod is dispersed and arranged in two or more positions in the fuel assembly, in order to achieve the same effect as in the first embodiment and to obtain a dynamic void coefficient greater than xe2x80x9cxe2x88x925¢xe2x80x9d, even though the width of the by-pass portion between the fuel assemblies is not changed. FIG. 5 is a sectional view showing the configuration of the fuel rods of the fuel assembly and the water-rods in this embodiment. As shown in FIG. 5, fuel assembly 36 bundles fuel rods 38 and six water-rods, and accommodates them in channel box 37 in a channel-shape. Boiling water stream path 41 is formed in this core. Two of the water-rods are water-rods 40a, 40b with a large diameter in the central section of the fuel assembly 36. Four other water-rods are provided as rods 39a, 39b, 39c, 39d arranged at four corner positions of the fuel assembly 36. Configuring water-rods at four corner positions of the fuel assembly 36 has the same effect as expanding d2 in the first embodiment. The absolute value of the void coefficient decreases when the amount of coolant increases. In the first embodiment, the by-pass portion is expanded to decrease the absolute value of the dynamic void coefficient. In this embodiment, by arranging water rods in the corners of the fuel assembly (adjacent to the by-pass portion 7) the same effect is achieved. Thus, a slowdown of neutrons can be promoted similar to expanding the width d2 of the by-pass portion 7. Therefore, like the first embodiment, the absolute value of the dynamic void coefficient can decrease, and xcex94MCPR of the pressure rise transient and a delay ratio of the in-core stability can be further improved. The magnitude of the cross-sectional area per one water-rod, the form, and the number of water-rods are not limited to the above-mentioned composition. The boiling water reactor of a fifth embodiment according to this invention is explained hereafter. In this embodiment, the dynamic void coefficient during rated operating power is like the dynamic void coefficient of the fourth embodiment, that is, greater than xe2x88x925¢ based on an average void fraction of 40%. In the fourth embodiment, four water-rods are arranged at four corner positions in the fuel assembly. The coolant (water) in these rods does not boil. But in this embodiment, the water-rods at four corner positions of this fuel assembly are eliminated, and the boiling water stream path is provided at the four corner positions. FIG. 6 is a sectional view showing the configuration of the fuel rods of the fuel assembly and the water-rods in this embodiment. As shown in FIG. 6, the fuel assembly 42 bundles two water-rods with a large diameter 40a and 40b located in the central part of the fuel assembly, and accommodates them in the channel box 37 having a channel-shape. The boiling water stream path 41 is formed. An area (symbol 43) where a fuel rod or a water-rod is not arranged is provided at the four corner positions of the fuel assembly 42. Thus, a fuel rod (of FIG. 13) is transposed to a moderator by deleting the fuel rods arranged at the four corner positions of the fuel assembly. Therefore, the same effect as in the first embodiment, where the width d2 of the by-pass portion 7 which serves as the coolant (i.e. non-boiling water) passage between the fuel assemblies, is achieved in this embodiment. Thus, like the fourth embodiment, the absolute value of the dynamic void coefficient is decreased, and as a result, the properties and the core stability during a pressure rise transient can be further improved. Furthermore, in this embodiment, the radius in the four vertex sections of the channel box can be enlarged as compared with the fuel assembly 36, or the conventional fuel assembly. That is, as compared with FIG. 5, four corner positions of the channel box 37 can be made roundish in FIG. 6. This reduces stress in such a channel box, and the thickness of the channel box 37 can be reduced. Furthermore, there is the advantage that the area of water is increased and the dynamic void coefficient is increased. The boiling water reactor of the sixth embodiment according to this invention is explained hereafter. In this embodiment, a mechanism involving a safety relief valve installed in the main steam pipe of the nuclear reactor is used. FIG. 7 shows the outline of such a boiling water reactor of this embodiment. For components of the same composition as in the conventional boiling water reactor, the same symbol as in FIG. 14 is used and additional explanation thereof is omitted. A relief safety valve 18xe2x80x2 and a turbine governor valve 19 are connected to the main steam pipe 17 connected to the pressure vessel 11. Usually, the relief safety valve 18 is in a closed state during operating power. Therefore, the steam is led to a high-pressure turbine 20 through main steam pipe 17 and turbine governor valve 19. A pressure rise transient occurs in a boiling water reactor, when turbine governor valve 19 is closed. In this embodiment, when the turbine governor valve 19 is closed, a relief safety valve opening signal 44 is inputted to at least one of the relief safety valves 18. By the input of the relief safety valve opening signal 44, the relief safety valve 18 is released, and the pressure in the pressure vessel will decline quickly. Therefore, since the increase in pressure in the pressure vessel at the time of a pressure rise transient can be controlled, xcex94MCPR is further improved with this action. The boiling water reactor of the seventh embodiment according to this invention is explained hereafter. In the sixth embodiment, an opening mechanism for the relief safety vale is employed. In this embodiment, equipment for forced circulation is provided and a part of total reactor core flow is provided by forced circulation. FIG. 8 shows the outline of such a boiling water reactor in this embodiment. For components of the same structure as in the conventional boiling water reactor of FIG. 14, the same symbol is used and additional explanation thereof is omitted. Hereinafter, only different parts are explained. In this embodiment, a circulating water-pump 45 is installed outside of the shroud 13 in the pressure vessel 11. A part of reactor core flow circulates by the circulating water pump 45 during operating power. When a pressure rise transient occurs, the turbine governor valve closes, and the circulating water pump stops (i.e. xe2x80x9cshuts downxe2x80x9d) due to a signal 46 inputted to the circulating water pump 45 from the valve. Even if the circulating water pump 45 stops, natural circulation maintains 70% or more of rated flow. As mentioned, due to the circulating water pump 45 stopping (or being partially stopped), when a part of flow is decreased at the time of a pressure rise transient, nuclear reactor power output will decline in a short time, and xcex94MCPR is further improved. In addition, according to this embodiment, since a part of reactor core flow is controllable by the forced circulation of the circulating water pump 45, power operation following the load on the power plant can be performed. When the circulating water pump 45 stops, it becomes impossible to ignore the temporary increase in xcex94MCPR by the flow transient. xcex94MCPR will increase significantly if the total forced flow ratio is larger than 30%. Therefore, in this embodiment, the forced-circulation flow rate is less than or equal to about 30% of total flow. Thus, even if the circulating water pump 45 stops, the system is designed such that 70% or more of the rated flow rate is maintained by natural circulation. The value of xcex94MCPR due to the reactor core flow ratio transient is thus always maintained smaller than xcex94MCPR due to the water-supply transient. Therefore, the reactor core flow ratio transient does not need to be taken into consideration in this embodiment, and xcex94MCPR does not get worse due to the reactor core flow ratio transient, like the conventional forced-circulation water reactor. The boiling water reactor of an eighth embodiment according to this invention is explained hereafter. In the sixth embodiment, an opening mechanism of a relief safety valve is employed. In this embodiment, a control rod drive which inserts a control rod from the upper part of the reactor core is formed in an upper part of the pressure vessel. FIG. 9 shows the outline of a boiling water reactor in this embodiment. For components of the same composition as in the conventional boiling water reactor of FIG. 14, the same symbol is used and additional explanation thereof is omitted. Hereinafter, only different parts are explained. In this embodiment, a control rod 15 is inserted from the upper part of reactor core 14, and a control rod drive 47 is connected with the control rod 15 and is installed above the upper part of the pressure vessel 11. A control rod guidance pipe 48 is provided to draw the control rod 15 from the reactor core 14 to the upper part of the shroud 13. When a pressure rise transient occurs, the control rod 15 falls in the reactor core 14 due to gravity, and controls reactor core reactivity quickly. The faster the velocity of the insertion (xe2x80x9cscramxe2x80x9d), the faster the power rise increase can be controlled and xcex94MCPR can be reduced further. Since gravity is used to insert the control rod 15, it is possible to perform reactivity control more rapidly than a system which inserts the control rod from the lower part, such as the one shown in FIG. 14. Furthermore, by adopting the top insertion method, it is possible to bring the position of the reactor core 14 close to the lower part of the pressure vessel 11, and as a result the center of gravity of the reactor core can be lowered, and earthquake-proof resistance can be increased. A boiling water reactor of the ninth embodiment according to this invention is explained hereafter. FIG. 10 is a graph which shows the progress of control rod density in this embodiment. Control rod density represents the portion of the number of control rod drives compared to the number of all control rod drives in core. In this embodiment, the control rod density is shown by the dash line (symbol 30a in the FIGURE) as burn-up progresses. Selected insertion of a part of a control rod is employed so that control rod density may be changed by about 10% at early stages of power operation (about 8 GWd/t). The control rods in the conventional nuclear reactor are taken all out in the last stage of the operating cycle. In this embodiment, in the last stage (about 8-12 GWd/t), although a few control rods are drawn out and control rod density is lowered from about 10% to about 1%, a state where the control rods are at least partially inserted is maintained. The solid line (symbol 30b in the FIGURE) shows the change of the excess reactivity accompanied by the change of the control rod density. This excess reactivity curve shows the reactor core reactivity under the assumption that all control rods are temporarily withdrawn. When at least a part of a control rod is always inserted into the reactor core (and rapid insertion of all control rods can be achieved), as compared with the case where all control rods are withdrawn, the input of reactivity is attained in a short time. Furthermore, xcex94MCPR is improved. In this embodiment, a change of the control rod density over the power operation cycle can be described as being a flat change. However, the same effect can be obtained not only by the above mentioned technique but also by a change which is not so flat. Moreover, this effect can be obtained if the control rods are inserted from the reactor core upper part or from the reactor core lower part. FIG. 11 is a sectional view showing the insertion position of the control rods of the reactor core in the last stage in this embodiment. This nuclear reactor core 49 shows one example where the fuel assembly 50 has 276 units. The symbol 52 represents a drawn-out state of a control rod and the symbol 51 represents an insertion state in the last stage of the operating cycle. In the last stage, 29 control rods among 61 control rods are inserted. FIG. 12 is a cross-section of the reactor core showing an example of the insertion state in the last stage of the operating cycle for a control rod 51. Here, a fuel assembly having the natural uranium blanket 34 is used at least for the lower edge. The embodiment of a fuel assembly having the natural uranium blanket areas 33 and 34 at both axial vertical edges is explained in connection with the third embodiment according to this invention and is also shown in FIG. 12. In this embodiment, the control rod 51 is inserted only to the lower blanket area 34 in the last stage of the operating cycle. FIG. 12 shows the situation where the maximum insertion is achieved, that is, it shows the embodiment where the control rod 51 is inserted to the upper edge of the lower blanket area 34. The length of the lower blanket area 34 is, for example, {fraction (3/24)} of the fuel effective section full length, and control rod density becomes about 5% when 29 control rods 51 (shown in FIG. 11) are inserted to the position of FIG. 12. Thus, the decline of reactor core reactivity can be minimized by inserting the control rods only in the blanket area of natural uranium in the lower part of the fuel assembly. Moreover, due to this insertion technique, since the reactivity can be controlled in a short time as compared with the case where all control rods are drawn out (like the conventional nuclear reactor), xcex94MCPR is further improved. Simultaneous rapid insertion of all the control rods during a scram is achieved by maintaining a state where a portion of the control rods are always inserted into the reactor core from the early stage to the last stage. In addition, in the above technique, although a change of the control rod density during operating power is completely flat as shown in FIG. 10, even if the density is not so flat, the same effect is ascertained. Moreover, the above-mentioned effect is the same even if the insertion direction of a control rod is from the reactor core lower or upper parts. The sixth to ninth embodiments may be used in a reactor having a dynamic void coefficient during rated operating power of greater than xe2x88x925¢ with an average void fraction of 40%, or with other reactors. According to this invention, xcex94MCPR during transients is minimized and the power density of the nuclear reactor is increased by reducing the delay ratio of the in-core stability. For these reasons, an economical natural circulation reactor (or a nuclear reactor using only partial forced-circulation) is achieved. The invention is, of course, not limited to the particular embodiments described above. Numerous variations and modifications of the above-described embodiments exist. One or more of the embodiments can be combined together. Accordingly, the invention is defined by the following claims. Japanese Priority Application Nos. PH10-241601, filed on Aug. 27, 1998, and PH11-181241, filed Jun. 28, 1999, including the specifications, drawings, claims and abstracts, are hereby incorporated by reference.