CA2724582A1 - Fuel for heavy water reactor or graphite reactor and process for producing the same - Google Patents
Fuel for heavy water reactor or graphite reactor and process for producing the same Download PDFInfo
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- CA2724582A1 CA2724582A1 CA2724582A CA2724582A CA2724582A1 CA 2724582 A1 CA2724582 A1 CA 2724582A1 CA 2724582 A CA2724582 A CA 2724582A CA 2724582 A CA2724582 A CA 2724582A CA 2724582 A1 CA2724582 A1 CA 2724582A1
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
- G21C3/42—Selection of substances for use as reactor fuel
- G21C3/58—Solid reactor fuel Pellets made of fissile material
- G21C3/62—Ceramic fuel
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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Abstract
A nuclear reactor fuel for use in a heavy water reactor or a graphite reactor that is produced by using a fuel material reprocessed from a spent fuel in a light water reactor without enriching U235 contained in the fuel material, and a process for producing a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor comprising the fuel material recovering step of recovering a fuel material from a spent fuel in a light water reactor, and the fuel producing step of producing the fuel by using the reprocessed fuel material without enriching U235 contained in the fuel material.
Description
r , FUEL FOR HEAVY WATER REACTOR OR GRAPHITE REACTOR
AND PROCESS FOR PRODUCING THE SAME
[Technical Field]
[0001]
The present invention relates to a fuel for a heavy water reactor or a graphite reactor and a process for producing the same.
In particular, it relates to a fuel for a heavy water reactor or a graphite reactor produced by using a fuel material reprocessed from a spent fuel in a light water reactor without enriching U235 contained in the fuel material, and a process for producing the same.
[Background Art]
AND PROCESS FOR PRODUCING THE SAME
[Technical Field]
[0001]
The present invention relates to a fuel for a heavy water reactor or a graphite reactor and a process for producing the same.
In particular, it relates to a fuel for a heavy water reactor or a graphite reactor produced by using a fuel material reprocessed from a spent fuel in a light water reactor without enriching U235 contained in the fuel material, and a process for producing the same.
[Background Art]
[0002]
In a nuclear reactor using light water as a moderator (hereinbelow referred to as a "light water reactor") that is widely used in Japan, a fuel in which U(uranium)235 is enriched up to about 3.6 to about 5% by mass (hereinbelow also referred to as "enriched uranium", and "% by mass" is simply referred to as "%") is used both in a PWR (pressurized water reactor) and a BWR
(boiling water reactor). Specifically, in the PWR, a fuel (STEP 1 fuel) having a degree of enrichment (hereinbelow simply referred to as an "enrichment") of U235 of 4.1% and a fuel (STEP 2 fuel) having the enrichment of U235 of 4.8% are used and, while in the BWR, a fuel having the enrichment of U235 of about 3.6% (in the case of a fuel assembly in which fuel rods are arranged in an 8 x 8 matrix), and a fuel having the enrichment of U235 of about 4% (in the case of a fuel assembly in which fuel rods are arranged in a 9 X 9 matrix) are used.
[00031 In each of the PWR and the BWR, when the fuel attains a given burnup, the fuel is discharged from the reactor as the spent fuel, and is subjected to reprocessing after being stored for a specific period of time. With regard to the specific given burnup in the PWR, the upper limit of the attained burnup (permissible maximum burnup) is 48 GWd/t for the fuel having the enrichment of 4.1%, and 55 GWd/t for the fuel having the enrichment of 4.8%.
However, the burnup differs depending on the position where the fuel is loaded in the reactor, and the fuel cannot be discharged from the reactor during the operation so that the fuels are actually discharged when the respective burnups are about 40 GWd/t and about 45 GWd/t, respectively. While in the BWR, the upper limit of the attained burnup is 48 GWd/t for the fuel having the enrichment of about 3.6% and 55 GWd/t for the fuel having the enrichment of about 4.0%, and the BWR has a reactor core larger than that of the PWR so that flexibility in disposition when the fuel is loaded in the reactor is high in the BWR, and hence the fuels are discharged at burnups close to the attained burnups mentioned above.
[00041 For reference purposes, the enrichment is set to the values mentioned above in consideration of that the cost for the enrichment is too high, and the reactivity of the reactor core immediately after new fuels are loaded is extremely increased (an infinite multiplication factor described later becomes extremely high) when the values are too high. In addition, the upper limits of the attained burnups are set to the values mentioned above in consideration of that U235 is reduced in response to the burnup and the number of nuclides absorbing neutrons generated by nuclear fission is increased so that it becomes difficult to maintain the reactor core in a critical state, and it is necessary to prevent damages to cladding tubes for a fuel pellet and a fuel rod.
[00051 The examination of the position of the fuel in the reactor, the loading of the fuel in the reactor, the determination of where or not the fuel has attained a given burnup, the discharging of the fuel from the reactor, and the storing of the fuel outside the reactor are actually performed on the basis of a fuel assembly obtained by assembling a large number of fuel rods into a given structure, a given shape, and given dimensions. However, this is an obvious matter, and the description becomes redundant when the matter is precisely described. Therefore, the fuel assembly will be simply referred to as a "fuel" unless particularly required to distinguish or refer to the fuel assembly.
[00061 Next, a description will be given for control over chain reaction of nuclear fission of U235 in a commercial reactor. In consideration of easiness in control, the chain reaction of the nuclear fission is continued by moderating, instead of using fast neutrons (having an energy of about 2 MeV) immediately after being generated by the nuclear fission, the fast neutrons into thermal neutrons (having an energy of about 0.0253 eV) by using a moderator, and then causing U235 to absorb the thermal neutrons. In addition, among neutrons generated by the nuclear fission of U 235, some are leaked out of the reactor during the moderation, and others are absorbed by the moderator, a neutron poison, U 238, or the nuclide generated by the nuclear fission.
The ratio of these neutrons that do not contribute to the chain reaction of the nuclear fission is increased as the burnup progresses. Accordingly, in the operation of an actual nuclear reactor, in order to compensate for a reduction in the reactivity of the reactor core in response to the burnup, various countermeasures are taken and various adjustments are made.
[0007]
For example, in the light water reactor, the concentration of the neutron poison in a coolant such as boron or the like is reduced in accordance with the progress in the burnup, the inserted portion of a control rod is reduced, the position where each fuel (assembly) to be burned in the subsequent operation cycle is loaded in the reactor is determined in consideration of its burnup after the completion of each operation cycle that continues for about one year, and a burnable poison is blended in the fuel, though the countermeasure and the adjustment of the PWR are slightly different from those of the BWR. Further, in a nuclear reactor using natural uranium or slightly enriched uranium, the reactor core thereof is larger than that of the light water reactor in general and the reactor core is not stored in a steel pressure vessel so that the fuel having attained the given burnup is discharged from the reactor during the operation, and a new fuel is loaded instead.
[0008]
In Japan, there are only light water reactors as commercial reactors. However, in the world, mainly in Canada, a nuclear reactor that uses natural uranium containing only 0.72% of U235 or slightly enriched uranium enriched to 1% to 2% as a fuel, 5 specifically, a reactor that uses heavy water as the moderator (e.g., a CANDU reactor), and a reactor that uses graphite as the moderator are operated as commercial reactors. The reason why such reactors are used is that a moderating ratio (average logarithmic energy decrement per collision x scattering cross section/absorption cross section) of the heavy water is 80 times that of the light water, the moderating ratio of graphite is 2.4 times that of the light water, and hence it becomes possible to maintain the reactor core in the critical state in the CANDU
reactor or the like using the heavy water even when natural uranium is used so that it becomes unnecessary to use expensive enriched uranium as the fuel.
[0009]
However, when the natural uranium or the slightly enriched uranium is used, because the content of U235 that undergoes nuclear fission is intrinsically small, the allowance of reactivity of the reactor core is rapidly reduced in response to the burnup in the reactor, and it is difficult to maintain the reactor core in the critical state for a long period of time. Consequently, the burnup where the fuel is to be discharged (discharge burnup) is much smaller than that in the light water reactor. For example, in totally 32 units of the CANDU reactor operated in 7 countries including Canada and Korea, the discharge burnup is about 7.5 GWd/t.
In a nuclear reactor using light water as a moderator (hereinbelow referred to as a "light water reactor") that is widely used in Japan, a fuel in which U(uranium)235 is enriched up to about 3.6 to about 5% by mass (hereinbelow also referred to as "enriched uranium", and "% by mass" is simply referred to as "%") is used both in a PWR (pressurized water reactor) and a BWR
(boiling water reactor). Specifically, in the PWR, a fuel (STEP 1 fuel) having a degree of enrichment (hereinbelow simply referred to as an "enrichment") of U235 of 4.1% and a fuel (STEP 2 fuel) having the enrichment of U235 of 4.8% are used and, while in the BWR, a fuel having the enrichment of U235 of about 3.6% (in the case of a fuel assembly in which fuel rods are arranged in an 8 x 8 matrix), and a fuel having the enrichment of U235 of about 4% (in the case of a fuel assembly in which fuel rods are arranged in a 9 X 9 matrix) are used.
[00031 In each of the PWR and the BWR, when the fuel attains a given burnup, the fuel is discharged from the reactor as the spent fuel, and is subjected to reprocessing after being stored for a specific period of time. With regard to the specific given burnup in the PWR, the upper limit of the attained burnup (permissible maximum burnup) is 48 GWd/t for the fuel having the enrichment of 4.1%, and 55 GWd/t for the fuel having the enrichment of 4.8%.
However, the burnup differs depending on the position where the fuel is loaded in the reactor, and the fuel cannot be discharged from the reactor during the operation so that the fuels are actually discharged when the respective burnups are about 40 GWd/t and about 45 GWd/t, respectively. While in the BWR, the upper limit of the attained burnup is 48 GWd/t for the fuel having the enrichment of about 3.6% and 55 GWd/t for the fuel having the enrichment of about 4.0%, and the BWR has a reactor core larger than that of the PWR so that flexibility in disposition when the fuel is loaded in the reactor is high in the BWR, and hence the fuels are discharged at burnups close to the attained burnups mentioned above.
[00041 For reference purposes, the enrichment is set to the values mentioned above in consideration of that the cost for the enrichment is too high, and the reactivity of the reactor core immediately after new fuels are loaded is extremely increased (an infinite multiplication factor described later becomes extremely high) when the values are too high. In addition, the upper limits of the attained burnups are set to the values mentioned above in consideration of that U235 is reduced in response to the burnup and the number of nuclides absorbing neutrons generated by nuclear fission is increased so that it becomes difficult to maintain the reactor core in a critical state, and it is necessary to prevent damages to cladding tubes for a fuel pellet and a fuel rod.
[00051 The examination of the position of the fuel in the reactor, the loading of the fuel in the reactor, the determination of where or not the fuel has attained a given burnup, the discharging of the fuel from the reactor, and the storing of the fuel outside the reactor are actually performed on the basis of a fuel assembly obtained by assembling a large number of fuel rods into a given structure, a given shape, and given dimensions. However, this is an obvious matter, and the description becomes redundant when the matter is precisely described. Therefore, the fuel assembly will be simply referred to as a "fuel" unless particularly required to distinguish or refer to the fuel assembly.
[00061 Next, a description will be given for control over chain reaction of nuclear fission of U235 in a commercial reactor. In consideration of easiness in control, the chain reaction of the nuclear fission is continued by moderating, instead of using fast neutrons (having an energy of about 2 MeV) immediately after being generated by the nuclear fission, the fast neutrons into thermal neutrons (having an energy of about 0.0253 eV) by using a moderator, and then causing U235 to absorb the thermal neutrons. In addition, among neutrons generated by the nuclear fission of U 235, some are leaked out of the reactor during the moderation, and others are absorbed by the moderator, a neutron poison, U 238, or the nuclide generated by the nuclear fission.
The ratio of these neutrons that do not contribute to the chain reaction of the nuclear fission is increased as the burnup progresses. Accordingly, in the operation of an actual nuclear reactor, in order to compensate for a reduction in the reactivity of the reactor core in response to the burnup, various countermeasures are taken and various adjustments are made.
[0007]
For example, in the light water reactor, the concentration of the neutron poison in a coolant such as boron or the like is reduced in accordance with the progress in the burnup, the inserted portion of a control rod is reduced, the position where each fuel (assembly) to be burned in the subsequent operation cycle is loaded in the reactor is determined in consideration of its burnup after the completion of each operation cycle that continues for about one year, and a burnable poison is blended in the fuel, though the countermeasure and the adjustment of the PWR are slightly different from those of the BWR. Further, in a nuclear reactor using natural uranium or slightly enriched uranium, the reactor core thereof is larger than that of the light water reactor in general and the reactor core is not stored in a steel pressure vessel so that the fuel having attained the given burnup is discharged from the reactor during the operation, and a new fuel is loaded instead.
[0008]
In Japan, there are only light water reactors as commercial reactors. However, in the world, mainly in Canada, a nuclear reactor that uses natural uranium containing only 0.72% of U235 or slightly enriched uranium enriched to 1% to 2% as a fuel, 5 specifically, a reactor that uses heavy water as the moderator (e.g., a CANDU reactor), and a reactor that uses graphite as the moderator are operated as commercial reactors. The reason why such reactors are used is that a moderating ratio (average logarithmic energy decrement per collision x scattering cross section/absorption cross section) of the heavy water is 80 times that of the light water, the moderating ratio of graphite is 2.4 times that of the light water, and hence it becomes possible to maintain the reactor core in the critical state in the CANDU
reactor or the like using the heavy water even when natural uranium is used so that it becomes unnecessary to use expensive enriched uranium as the fuel.
[0009]
However, when the natural uranium or the slightly enriched uranium is used, because the content of U235 that undergoes nuclear fission is intrinsically small, the allowance of reactivity of the reactor core is rapidly reduced in response to the burnup in the reactor, and it is difficult to maintain the reactor core in the critical state for a long period of time. Consequently, the burnup where the fuel is to be discharged (discharge burnup) is much smaller than that in the light water reactor. For example, in totally 32 units of the CANDU reactor operated in 7 countries including Canada and Korea, the discharge burnup is about 7.5 GWd/t.
[00101 It is to be noted that, in the nuclear reactor using the natural uranium or the slightly enriched uranium, e.g., the CANDU reactor using the heavy water, a technology in which enriched uranium is loaded to increase the discharge burnup, which is used in the light water reactor, is not put into practical use. The reasons for the above are that enriched uranium is expensive, and that the reactivity of the nuclear reactor, particularly the reactivity immediately after the loading, is extremely increased so that it becomes difficult to operate the reactor safely when the enriched uranium without being processed is used. As the countermeasure against that, when a large number of neutron poisons is blended into the enriched uranium to lower the reactivity, neutron economy deteriorates.
[00111 Subsequently, both in the light water reactor and the nuclear reactor using the natural uranium or the slightly enriched uranium, the fuel discharged from the reactor as the spent fuel, the fuel having attained the given burnup, is stored for a specific period of time in a storage pool having specific equipment because intense radiation is emitted from radioactive elements (daughter nuclides) generated by the nuclear fission of U235 and heat is generated. The fuel is then subjected to reprocessing such as recovering unburned uranium, recovering plutonium generated by the burnup of uranium, and the like (Non-Patent Document 1).
[Non-Patent Document 11 R. Kiyose, "Chemical Engineering of spent fuel and plutonium", Nikkan Kogyo Shimbun Ltd., 1984 [Disclosure of the Invention]
[Problem to be Solved by the Invention]
[0012]
However, with a steep rise in oil price in recent years, a reduction in fuel cycle cost, an improvement in resource utilization rate (i.e. effective utilization of resources), and a reduction in radioactive waste have been increasingly demanded both in the nuclear reactor using the natural uranium or the slightly enriched uranium and the light water reactor.
[0013]
Accordingly, not only in the nuclear reactor using the natural uranium or the slightly enriched uranium but also in the light water reactor, the development of the technology capable of the reduction in fuel cycle cost, the improvement in resource utilization rate, and the reduction in radioactive waste has been demanded.
[Means for Solving the Problem]
[0014]
The present invention has been achieved in order to solve the foregoing problems, and make the spent fuel in a light water reactor capable of being burned in a nuclear reactor using natural uranium or slightly enriched uranium. A description will be given hereinbelow for the invention of each claim.
[0015]
The invention according to claim 1 is a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor that is produced by using a fuel material reprocessed from a spent fuel in a light water reactor without enriching U235 contained in the fuel material.
[0016]
In the invention according to claim 1, because the fuel material reprocessed from the spent fuel in the light water reactor is reburned in the heavy water reactor or the graphite reactor that normally uses natural uranium or slightly enriched uranium as the fuel without enriching U235 contained in the fuel material, it is possible to achieve a reduction in fuel cycle cost, an improvement in resource utilization rate, and a reduction in radioactive waste in the heavy water reactor or the graphite reactor. On the other hand, in the light water reactor as well, it is possible to achieve the reduction in fuel cycle cost resulting from a reduction in enrichment cost and elimination of an enrichment process, effective utilization of resources, and the reduction in radioactive waste.
[0017]
The "spent fuel" mentioned above includes not only a fuel that has attained the upper limit of an attained burnup or a burnup close to the upper limit thereof but also a fuel that, even when trying to burn the fuel again in the subsequent cycle, will attain the permissible maximum burnup during the subsequent cycle operation so that it may not be burned in the subsequent cycle operation as a result.
[0018]
In the currently used light water reactor, though a uranium fuel obtained by enriching natural uranium is mainly used, it is planned to use a MOX fuel blended with plutonium in the future.
In addition, a thorium fuel will be developed in the future.
[00111 Subsequently, both in the light water reactor and the nuclear reactor using the natural uranium or the slightly enriched uranium, the fuel discharged from the reactor as the spent fuel, the fuel having attained the given burnup, is stored for a specific period of time in a storage pool having specific equipment because intense radiation is emitted from radioactive elements (daughter nuclides) generated by the nuclear fission of U235 and heat is generated. The fuel is then subjected to reprocessing such as recovering unburned uranium, recovering plutonium generated by the burnup of uranium, and the like (Non-Patent Document 1).
[Non-Patent Document 11 R. Kiyose, "Chemical Engineering of spent fuel and plutonium", Nikkan Kogyo Shimbun Ltd., 1984 [Disclosure of the Invention]
[Problem to be Solved by the Invention]
[0012]
However, with a steep rise in oil price in recent years, a reduction in fuel cycle cost, an improvement in resource utilization rate (i.e. effective utilization of resources), and a reduction in radioactive waste have been increasingly demanded both in the nuclear reactor using the natural uranium or the slightly enriched uranium and the light water reactor.
[0013]
Accordingly, not only in the nuclear reactor using the natural uranium or the slightly enriched uranium but also in the light water reactor, the development of the technology capable of the reduction in fuel cycle cost, the improvement in resource utilization rate, and the reduction in radioactive waste has been demanded.
[Means for Solving the Problem]
[0014]
The present invention has been achieved in order to solve the foregoing problems, and make the spent fuel in a light water reactor capable of being burned in a nuclear reactor using natural uranium or slightly enriched uranium. A description will be given hereinbelow for the invention of each claim.
[0015]
The invention according to claim 1 is a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor that is produced by using a fuel material reprocessed from a spent fuel in a light water reactor without enriching U235 contained in the fuel material.
[0016]
In the invention according to claim 1, because the fuel material reprocessed from the spent fuel in the light water reactor is reburned in the heavy water reactor or the graphite reactor that normally uses natural uranium or slightly enriched uranium as the fuel without enriching U235 contained in the fuel material, it is possible to achieve a reduction in fuel cycle cost, an improvement in resource utilization rate, and a reduction in radioactive waste in the heavy water reactor or the graphite reactor. On the other hand, in the light water reactor as well, it is possible to achieve the reduction in fuel cycle cost resulting from a reduction in enrichment cost and elimination of an enrichment process, effective utilization of resources, and the reduction in radioactive waste.
[0017]
The "spent fuel" mentioned above includes not only a fuel that has attained the upper limit of an attained burnup or a burnup close to the upper limit thereof but also a fuel that, even when trying to burn the fuel again in the subsequent cycle, will attain the permissible maximum burnup during the subsequent cycle operation so that it may not be burned in the subsequent cycle operation as a result.
[0018]
In the currently used light water reactor, though a uranium fuel obtained by enriching natural uranium is mainly used, it is planned to use a MOX fuel blended with plutonium in the future.
In addition, a thorium fuel will be developed in the future.
Consequently, the "fuel" mentioned above is not limited to the uranium fuel.
[0019]
Further, the "fuel material" mentioned above denotes a fissile material and its isotope. U235, Pu239 and isotopes thereof are currently used in light water reactors in Japan.
However, thorium, which may be used in the future, is not excluded and other impurity elements that can be inevitably contained may also be included.
[0020]
Furthermore, the description "produced by using a fuel material without enriching U235 contained in the fuel material"
mentioned above includes the case where production is performed by using even Pu239 or the like as the fissile material without separating uranium and plutonium from the spent fuel as long as U235 in uranium reprocessed from the spent fuel is not enriched.
The case where production is performed by further adding a fissile material reprocessed from a nuclear bomb, natural uranium, or uranium obtained by slightly enriching natural uranium is also included, in this case. Moreover, the case where production is performed by removing uranium from the reprocessed fuel material and using only plutonium is also included.
[0021]
Moreover, the description "produced without enriching U235" mentioned above means that the enrichment of U235 in the reprocessed uranium is not performed when the fuel for the heavy water reactor or the graphite reactor is produced (unlike the natural uranium, the reprocessed uranium contains U232 functioning as an intense radiation source so that an enrichment process becomes complicated and costly).
[0022]
Fuel pellets in fuel rods constituting a nuclear reactor fuel 5 assembly also correspond to the "nuclear reactor fuel" of the present invention and, even when the number of fuel pellets is one, the fuel pellet corresponds to the "nuclear reactor fuel" of the present invention.
[0023]
[0019]
Further, the "fuel material" mentioned above denotes a fissile material and its isotope. U235, Pu239 and isotopes thereof are currently used in light water reactors in Japan.
However, thorium, which may be used in the future, is not excluded and other impurity elements that can be inevitably contained may also be included.
[0020]
Furthermore, the description "produced by using a fuel material without enriching U235 contained in the fuel material"
mentioned above includes the case where production is performed by using even Pu239 or the like as the fissile material without separating uranium and plutonium from the spent fuel as long as U235 in uranium reprocessed from the spent fuel is not enriched.
The case where production is performed by further adding a fissile material reprocessed from a nuclear bomb, natural uranium, or uranium obtained by slightly enriching natural uranium is also included, in this case. Moreover, the case where production is performed by removing uranium from the reprocessed fuel material and using only plutonium is also included.
[0021]
Moreover, the description "produced without enriching U235" mentioned above means that the enrichment of U235 in the reprocessed uranium is not performed when the fuel for the heavy water reactor or the graphite reactor is produced (unlike the natural uranium, the reprocessed uranium contains U232 functioning as an intense radiation source so that an enrichment process becomes complicated and costly).
[0022]
Fuel pellets in fuel rods constituting a nuclear reactor fuel 5 assembly also correspond to the "nuclear reactor fuel" of the present invention and, even when the number of fuel pellets is one, the fuel pellet corresponds to the "nuclear reactor fuel" of the present invention.
[0023]
10 It is to be noted that the "heavy water reactor" and the "graphite reactor" include a nuclear reactor using a gas or light water as a coolant as long as the nuclear reactor uses the heavy water or the graphite as the moderator.
[0024]
The invention according to claim 2 is the nuclear reactor fuel according to claim 1, wherein the fuel material reprocessed from the spent fuel in the light water reactor is uranium.
[0025]
In this invention of claim 2, since the nuclear reactor fuel for use in the heavy water reactor or the graphite reactor is produced by using uranium reprocessed from the spent fuel in the light water reactor, mainly U235 left unburned in the light water reactor, and a minute amount of other uranium isotopes are burned in the heavy water reactor or the graphite reactor, which leads to a reduction in fuel cost not only in the heavy water reactor or the graphite reactor but also in the light water reactor.
[0026]
In addition, the spent fuel in the light water reactor contains more U235 as the fissile material than the natural uranium, and hence it is possible to especially favorably exert the effect of the invention of claim 1.
[0027]
The invention according to claim 3 is the nuclear reactor fuel according to claim 2, wherein the light water reactor is a PWR.
[0028]
In the invention of claim 3, it is possible to exert the effect of the invention of claim 1 or claim 2 further favorably. The reason for that is shown below. The PWR has a reactor core smaller than that of a BWR, and flexibility in disposing the fuel in the reactor is low so that a large amount of the fuel that has to be determined as the spent fuel occurs well before the permissible maximum burnup is attained. As a result, a composition ratio of U235 in the fuel determined as the spent fuel is higher than the composition ratio thereof in the natural uranium, and it is possible to increase the burnup and prolong the interval of the fuel exchange correspondingly when the fuel is used in the heavy water reactor or the graphite reactor.
[0029]
The invention according to claim 4 is the nuclear reactor fuel according to claim 1 or claim 2, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
[0030]
The invention according to claim 5 is the nuclear reactor fuel according to claim 3, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
[0024]
The invention according to claim 2 is the nuclear reactor fuel according to claim 1, wherein the fuel material reprocessed from the spent fuel in the light water reactor is uranium.
[0025]
In this invention of claim 2, since the nuclear reactor fuel for use in the heavy water reactor or the graphite reactor is produced by using uranium reprocessed from the spent fuel in the light water reactor, mainly U235 left unburned in the light water reactor, and a minute amount of other uranium isotopes are burned in the heavy water reactor or the graphite reactor, which leads to a reduction in fuel cost not only in the heavy water reactor or the graphite reactor but also in the light water reactor.
[0026]
In addition, the spent fuel in the light water reactor contains more U235 as the fissile material than the natural uranium, and hence it is possible to especially favorably exert the effect of the invention of claim 1.
[0027]
The invention according to claim 3 is the nuclear reactor fuel according to claim 2, wherein the light water reactor is a PWR.
[0028]
In the invention of claim 3, it is possible to exert the effect of the invention of claim 1 or claim 2 further favorably. The reason for that is shown below. The PWR has a reactor core smaller than that of a BWR, and flexibility in disposing the fuel in the reactor is low so that a large amount of the fuel that has to be determined as the spent fuel occurs well before the permissible maximum burnup is attained. As a result, a composition ratio of U235 in the fuel determined as the spent fuel is higher than the composition ratio thereof in the natural uranium, and it is possible to increase the burnup and prolong the interval of the fuel exchange correspondingly when the fuel is used in the heavy water reactor or the graphite reactor.
[0029]
The invention according to claim 4 is the nuclear reactor fuel according to claim 1 or claim 2, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
[0030]
The invention according to claim 5 is the nuclear reactor fuel according to claim 3, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
[0031]
Among the inventions according to the above claims, it is possible to exert the effect of the inventions according to any one of claims 1-3 most favorably. The reason for the above is shown below. That is, because the CANDU reactor uses the heavy water having high moderating ratio as the moderator, even when the composition ratio of the fissile material in the fuel is significantly reduced, it is possible to maintain the reactor core in the critical state. Consequently, the fuel for the CANDU reactor, which is produced by using the fuel material reprocessed from the fuel having attained the maximum burnup in the light water reactor, is capable of obtaining a sufficiently high burnup as compared with the fuel using the natural uranium in many cases.
[0032]
The invention according to claim 6 is a process for producing a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor comprising the fuel material recovering step of recovering a fuel material from a spent fuel in a light water reactor, and the fuel producing step of producing the fuel by using the reprocessed fuel material without enriching U235 contained in the fuel material.
[0033]
The invention according to claim 6 is an aspect of the present invention obtained by viewing the invention according to claim 1 as a process.
[0034]
The invention according to claim 7 is the process for producing a nuclear reactor fuel according to claim 6 further comprising, after the fuel material recovering step, the composition ratio calculating step of calculating a composition ratio of each composition material in the reprocessed fuel material, the value coefficient calculating step of calculating a quotient by dividing a product of a fission cross section and a number of neutrons generated by nuclear fission by captured cross section for each composition material to determine the quotient as a value coefficient for each composition material, the value calculating step of calculating a product of the composition ratio and the value coefficient for each composition material to determine a sum total of the product as a value of the reprocessed fuel material, and the reuse determining step of comparing the value of the fuel material with a given reference value to determine whether or not the fuel producing step is to be performed according to a magnitude relationship therebetween.
[00351 In the invention according to claim 7, since the value of the reprocessed fuel material is compared with the given reference value to determine whether or not the fuel material is to be used as the fuel for the heavy water reactor or the graphite reactor, when the fuel produced by using the spent fuel in the light water reactor is used in the heavy water reactor or the graphite reactor as the result of the determination, it is possible to increase the attained burnup and prolong the interval of the fuel exchange.
[00361 In the strict sense, although the fission cross section (of), the number of neutrons generated by the nuclear fission (v), and a absorption cross section (oa) are dependent on neutron energy, in the present invention, because the target reactor is a thermal neutron reactor such as the CANDU reactor or the like, and most of the neutrons contributing to the nuclear reaction in the reactor are thermal neutrons, the value coefficient is calculated using values in the thermal neutron energy (0.0253 eV) by the following expression:
value coefficient = v X oa/of.
[00371 The "composition ratio of each composition material in the reprocessed fuel material" may be a composition ratio between a uranium isotope and a plutonium isotope in a uranium and plutonium mixture, or the composition ratios for each uranium and for each plutonium. Further, the composition ratio of another atom or element that is inevitably generated or mixed into the fuel material may also be included. A means for determining the composition ratio may be any means such as calculation, actual measurement, an experience, or the like.
[00381 Furthermore, as the "given reference value", there may be adopted a value determined by the same procedure as the value calculating step for the fuel material in claim 7 on the basis of the composition ratio of each composition material, the fission cross section, the number of neutrons generated by the nuclear fission, and the absorption cross section of a fuel material that is normally burned in the reactor in which the fuel material of concern is burned, e.g., the natural uranium in the case of the CANDU reactor. There may be adopted a value obtained by modifying the above-mentioned value in which differences in the type of the reactor and neutron spectrum, an error in the measurement of the composition ratio, and a cost required to use the fuel material as the fuel for the CANDU reactor, for example, are reflected.
5 [0039]
Moreover, when plutonium or U235 discharged from a nuclear bomb is blended into the "reprocessed fuel material", naturally, the given reference value is prepared in consideration of the blended plutonium or U235. In this case, the value of the 10 fuel material including the "reprocessed fuel material" and the "blended plutonium or U235 discharged from the nuclear bomb"
may be calculated and compared with the given reference value (which is equivalent to the present invention, and such case is included in the invention of claim 7). In addition, the same 15 applies to the case where U235 that is slightly enriched separately is blended.
[0040]
Further, a plurality of reference values for the heavy water reactor and the graphite reactor may be prepared, and the reactor to be reused may be changed in accordance with the values.
[0041]
The invention according to claim 8 is a process for producing a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor comprising the uranium recovering step of recovering uranium from a spent fuel in a light water reactor, and the fuel producing step of producing the fuel by using the reprocessed uranium without enriching U235 contained in the reprocessed uranium as a composition material.
Among the inventions according to the above claims, it is possible to exert the effect of the inventions according to any one of claims 1-3 most favorably. The reason for the above is shown below. That is, because the CANDU reactor uses the heavy water having high moderating ratio as the moderator, even when the composition ratio of the fissile material in the fuel is significantly reduced, it is possible to maintain the reactor core in the critical state. Consequently, the fuel for the CANDU reactor, which is produced by using the fuel material reprocessed from the fuel having attained the maximum burnup in the light water reactor, is capable of obtaining a sufficiently high burnup as compared with the fuel using the natural uranium in many cases.
[0032]
The invention according to claim 6 is a process for producing a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor comprising the fuel material recovering step of recovering a fuel material from a spent fuel in a light water reactor, and the fuel producing step of producing the fuel by using the reprocessed fuel material without enriching U235 contained in the fuel material.
[0033]
The invention according to claim 6 is an aspect of the present invention obtained by viewing the invention according to claim 1 as a process.
[0034]
The invention according to claim 7 is the process for producing a nuclear reactor fuel according to claim 6 further comprising, after the fuel material recovering step, the composition ratio calculating step of calculating a composition ratio of each composition material in the reprocessed fuel material, the value coefficient calculating step of calculating a quotient by dividing a product of a fission cross section and a number of neutrons generated by nuclear fission by captured cross section for each composition material to determine the quotient as a value coefficient for each composition material, the value calculating step of calculating a product of the composition ratio and the value coefficient for each composition material to determine a sum total of the product as a value of the reprocessed fuel material, and the reuse determining step of comparing the value of the fuel material with a given reference value to determine whether or not the fuel producing step is to be performed according to a magnitude relationship therebetween.
[00351 In the invention according to claim 7, since the value of the reprocessed fuel material is compared with the given reference value to determine whether or not the fuel material is to be used as the fuel for the heavy water reactor or the graphite reactor, when the fuel produced by using the spent fuel in the light water reactor is used in the heavy water reactor or the graphite reactor as the result of the determination, it is possible to increase the attained burnup and prolong the interval of the fuel exchange.
[00361 In the strict sense, although the fission cross section (of), the number of neutrons generated by the nuclear fission (v), and a absorption cross section (oa) are dependent on neutron energy, in the present invention, because the target reactor is a thermal neutron reactor such as the CANDU reactor or the like, and most of the neutrons contributing to the nuclear reaction in the reactor are thermal neutrons, the value coefficient is calculated using values in the thermal neutron energy (0.0253 eV) by the following expression:
value coefficient = v X oa/of.
[00371 The "composition ratio of each composition material in the reprocessed fuel material" may be a composition ratio between a uranium isotope and a plutonium isotope in a uranium and plutonium mixture, or the composition ratios for each uranium and for each plutonium. Further, the composition ratio of another atom or element that is inevitably generated or mixed into the fuel material may also be included. A means for determining the composition ratio may be any means such as calculation, actual measurement, an experience, or the like.
[00381 Furthermore, as the "given reference value", there may be adopted a value determined by the same procedure as the value calculating step for the fuel material in claim 7 on the basis of the composition ratio of each composition material, the fission cross section, the number of neutrons generated by the nuclear fission, and the absorption cross section of a fuel material that is normally burned in the reactor in which the fuel material of concern is burned, e.g., the natural uranium in the case of the CANDU reactor. There may be adopted a value obtained by modifying the above-mentioned value in which differences in the type of the reactor and neutron spectrum, an error in the measurement of the composition ratio, and a cost required to use the fuel material as the fuel for the CANDU reactor, for example, are reflected.
5 [0039]
Moreover, when plutonium or U235 discharged from a nuclear bomb is blended into the "reprocessed fuel material", naturally, the given reference value is prepared in consideration of the blended plutonium or U235. In this case, the value of the 10 fuel material including the "reprocessed fuel material" and the "blended plutonium or U235 discharged from the nuclear bomb"
may be calculated and compared with the given reference value (which is equivalent to the present invention, and such case is included in the invention of claim 7). In addition, the same 15 applies to the case where U235 that is slightly enriched separately is blended.
[0040]
Further, a plurality of reference values for the heavy water reactor and the graphite reactor may be prepared, and the reactor to be reused may be changed in accordance with the values.
[0041]
The invention according to claim 8 is a process for producing a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor comprising the uranium recovering step of recovering uranium from a spent fuel in a light water reactor, and the fuel producing step of producing the fuel by using the reprocessed uranium without enriching U235 contained in the reprocessed uranium as a composition material.
[0042]
The invention according to claim 8 is an aspect of the present invention obtained by viewing the invention according to claim 2 as a process.
[0043]
The invention according to claim 9 is the process for producing a nuclear reactor fuel of the invention according to claim 8 comprising, after the uranium recovering step, the composition ratio calculating step of calculating a composition ratio of each uranium isotope in the reprocessed uranium, the value coefficient calculating step of calculating a quotient by dividing a product of a fission cross section and a number of neutrons generated by nuclear fission by a absorption cross section for each uranium isotope to determine the quotient as a value coefficient of each uranium isotope, the value calculating step of calculating a product of the composition ratio and the value coefficient for each uranium isotope to determine a sum total of the product as a value of the reprocessed uranium, and the reuse determining step of comparing the value of the reprocessed uranium with a given reference value to determine whether or not the fuel producing step is to be performed according to a magnitude relationship therebetween.
[0044]
In the invention of claim 9, since the value of the reprocessed uranium is compared with the given reference value to determine whether or not the fuel material is to be used as the fuel for the heavy water reactor or the graphite reactor, when the fuel produced by using uranium reprocessed from the spent fuel in the light water reactor is used in the heavy water reactor or the graphite reactor, it is possible to increase the attained burnup and prolong the interval of the fuel exchange.
[00451 It is to be noted that the "uranium" mentioned above includes uranium in which other elements are inevitably contained. Consequently, when the value of the reprocessed uranium is determined, the value of the reprocessed uranium may be modified in consideration of influence exerted by such impurity elements on the basis of an experience or actual measurement.
[00461 The invention of claim 10 is the process for producing a nuclear reactor fuel according to any one of claims 6-9, wherein the light water reactor is a PWR.
[00471 In the invention of claim 10, it is possible to exert the effect of the invention according to any one of claims 6-9 further favorably.
[00481 The invention according to claim 11 is the process for producing a nuclear reactor fuel according to any one of claims 6-9, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
[00491 In addition, the invention according to claim 12 is the process for producing a nuclear reactor fuel according to claim 10, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
The invention according to claim 8 is an aspect of the present invention obtained by viewing the invention according to claim 2 as a process.
[0043]
The invention according to claim 9 is the process for producing a nuclear reactor fuel of the invention according to claim 8 comprising, after the uranium recovering step, the composition ratio calculating step of calculating a composition ratio of each uranium isotope in the reprocessed uranium, the value coefficient calculating step of calculating a quotient by dividing a product of a fission cross section and a number of neutrons generated by nuclear fission by a absorption cross section for each uranium isotope to determine the quotient as a value coefficient of each uranium isotope, the value calculating step of calculating a product of the composition ratio and the value coefficient for each uranium isotope to determine a sum total of the product as a value of the reprocessed uranium, and the reuse determining step of comparing the value of the reprocessed uranium with a given reference value to determine whether or not the fuel producing step is to be performed according to a magnitude relationship therebetween.
[0044]
In the invention of claim 9, since the value of the reprocessed uranium is compared with the given reference value to determine whether or not the fuel material is to be used as the fuel for the heavy water reactor or the graphite reactor, when the fuel produced by using uranium reprocessed from the spent fuel in the light water reactor is used in the heavy water reactor or the graphite reactor, it is possible to increase the attained burnup and prolong the interval of the fuel exchange.
[00451 It is to be noted that the "uranium" mentioned above includes uranium in which other elements are inevitably contained. Consequently, when the value of the reprocessed uranium is determined, the value of the reprocessed uranium may be modified in consideration of influence exerted by such impurity elements on the basis of an experience or actual measurement.
[00461 The invention of claim 10 is the process for producing a nuclear reactor fuel according to any one of claims 6-9, wherein the light water reactor is a PWR.
[00471 In the invention of claim 10, it is possible to exert the effect of the invention according to any one of claims 6-9 further favorably.
[00481 The invention according to claim 11 is the process for producing a nuclear reactor fuel according to any one of claims 6-9, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
[00491 In addition, the invention according to claim 12 is the process for producing a nuclear reactor fuel according to claim 10, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
[0050]
Among the inventions according to the claims, it is possible to exert the effect of inventions according to any one of claims 6-10 most favorably.
[Effect of the Invention]
[0051]
In the present invention, since a spent fuel in a light water reactor is burned as the fuel for a heavy water reactor or a graphite reactor, it is possible to achieve a reduction in fuel cycle cost, an improvement in resource utilization rate, and a reduction in radioactive waste not only in a nuclear reactor using natural uranium or slightly enriched uranium but also in the light water reactor.
[Brief Description of the Drawings]
[0052]
FIG. 1 is a view conceptually showing an outer appearance of a fuel assembly for a CANDU reactor;
FIG. 2 is a view showing neutron flux spectrums (neutron energy distributions) in the CANDU reactor and a PWR;
FIG. 3 is a view showing changes in infinite multiplication factor in response to burnup of natural uranium and reprocessed uranium in the CANDU reactor;
FIG. 4 is a view conceptually showing changes in burnup and infinite multiplication factor when 6 types of reprocessed uranium reprocessed from fuels burned in the PWR and natural uranium are burned in the CANDU reactor; and FIG. 5 is a view conceptually showing changes in burnup and infinite multiplication factor when 6 types of reprocessed uranium reprocessed from fuels burned in a BWR and natural uranium are burned in the CANDU reactor.
[Explanations of Letters or Numerals]
[0053]
11 bearing pad 12 cladding tube 13 end support plate 14 fuel pellet spacer 16 end cap 15 [Best Mode for Carrying Out the Invention]
[0054]
A description will be given hereinbelow of embodiments of the present invention with reference to drawings. It is to be noted that the present invention is not limited to the following embodiments. Various modifications may be made to the following embodiments within the range identical or equivalent to the present invention.
[0055]
(With regard to use of reprocessed uranium in CANDU reactor) The present first embodiment relates to burning, in a CANDU reactor, of uranium reprocessed from a spent fuel in a PWR instead of natural uranium. A description will be given hereinbelow with reference to drawings and tables.
[0056]
In Table 1, there are shown a composition (%) of isotopes of uranium reprocessed from a fuel (uranium fuel) that has been discharged from a reactor as a spent fuel because the fuel has 5 attained a given burnup in a PWR having an electrical power of 1000 MWe, and has been left for 150 days, i.e., reprocessed uranium, together with a composition of isotopes of natural uranium.
[0057]
10 [Table 1]
uranium isotope natural uranium reprocessed uranium U234 0.0055 0.0121 U235 0.72 0.8259 U236 0 0.4379 U238 99.2745 98.7241 Total 100 100 [0058]
In Table 2, there are shown a fission cross section (of) and a absorption cross section (oa) of each of uranium isotopes in 15 thermal neutron energy (0.0253 eV) (Japanese Evaluated Nuclear Data Library JENDL-3.3, Japan Atomic Energy Agency, July 27, 2007). It is to be noted that the unit is barn.
[0059]
[Table 2]
uranium isotope fission cross section absorption cross section Of Ga U234 2.95x10-1 9.975x 101 U235 5.851x10-2 9.869x101 U236 6.129x 10.2 5.295 U238 1.177x105 2.717 [0060]
A fuel assembly for the CANDU reactor shown in FIG. 1 was designed using the natural uranium and the reprocessed uranium each having the composition shown in Table 1. In FIG. 1, reference numeral 11 denotes a bearing pad, reference numeral 12 denotes a cladding tube, reference numeral 13 denotes an end support plate, reference numeral 14 denotes a fuel pellet loaded in the cladding tube 12, reference numeral 15 denotes a spacer, and reference numeral 16 denotes an end cap. The end support plate 13 is fixed to an end portion of a fuel assembly 10 by resistance welding, the end cap 16 is fixed by resistance welding in order to seal an end portion of the cladding tube 12, and the spacer 15 and the bearing pad 11 are soldered to the cladding tube 12.
[0061]
The length of the fuel assembly 10 is 49.5 cm, the outer diameter of the cladding tube 12 is 13.061 mm, the wall thickness thereof is 0.419 mm, the diameter of the fuel pellet 15 is 12.154 mm, the length thereof is 16.40 mm, and the density thereof is 10.6 g/cm 3 .
[0062]
Next, for different fuels (the natural uranium and the reprocessed uranium) of the fuel assembly, neutron energy distributions (neutron flux spectrums) in the fuels were calculated. The result is shown in FIG. 2. In FIG. 2, the horizontal axis is a logarithmic scale of neutron energy (eV), while the vertical axis indicates a neutron density (Source/Lethargy/Volume) in the fuel. A thick line indicates the neutron flux spectrum of the natural uranium, while a thin line indicates that of the reprocessed uranium. The neutron flux spectrum in the PWR is also shown in FIG. 2 for reference purposes.
[0063]
In FIG. 2, the thick line can be barely distinguished from the thin line in a region from about 0.3 x 1E - 2 eV to about 0.3 X
1E - 1 eV, but they completely conform to each other in other regions. That is, the neutron spectrums of the natural uranium and the reprocessed uranium in the CANDU reactor are scarcely different from each other in all neutron energy regions, and it can be seen from the result that it is possible to burn the reprocessed uranium without being processed in the CANDU reactor. It is inferred that the reason why the neutron flux is larger in the CANDU reactor than in the PWR in the region where the energy is from 0.3 x 1E - 2 eV to 0.3 x 1E - 1 eV is that the CANDU reactor has a large reactor core so that the leakage of the neutron is small, and the moderating ratio of the heavy water as the moderator is large.
[00641 Subsequently, the calculation of burnup of the respective fuels was performed to compare changes in infinite multiplication factor in response to the burnup. The result is shown in FIG. 3.
In FIG. 3, the horizontal axis indicates the burnup (the burnup of the fuel assembly) (MWd/t), the vertical axis on the left indicates the infinite multiplication factor (k-infinity), and the vertical axis on the right indicates a difference between infinite multiplication factors of both fuels Ak (the infinite multiplication factor of the reprocessed uranium - that of the natural uranium) W. In the drawing, a line joining black diamonds indicates a natural uranium fuel, a line joining black squares indicates a reprocessed uranium fuel, and a line joining open circles indicates the difference between infinite multiplication factors of both fuels.
[00651 Although the currently used CANDU reactor adopts 7500 MWd/t as a discharge burnup, the discharge burnup of the reprocessed uranium corresponding to the infinite multiplication factor at this point is about 9500 MWd/t, as indicated by arrows in FIG. 3. This denotes that, when the reprocessed uranium fuel is used in the CANDU reactor, the fuel can be used for a period prolonged by about 25%, and it becomes possible to achieve a reduction in fuel cycle cost and a reduction in radioactive waste.
[00661 (With regard to value of reprocessed uranium) Next, a description will be given hereinbelow of a value coefficient of the uranium isotope and a value of the reprocessed uranium.
In Table 1, more U235 that is liable to undergo the nuclear fission reaction is contained in the reprocessed uranium than in the natural uranium and, as a result, it can be said that the reprocessed uranium is a fuel having a value higher than that of the natural uranium. However, the composition of the uranium isotopes in the uranium reprocessed from the light water reactor does not always conform to the composition in Table 1, and is dependent on the enrichment when loaded in the light water reactor, and the discharge burnup when discharged from the light water reactor.
[00671 As shown in Table 1, U234 of which only a minute amount is contained in the natural uranium and U236 that does not exist in the natural uranium are produced and contained in the reprocessed uranium, and they function as poisons for capturing neutrons. Consequently, for example, when the proportion of U235 in the reprocessed uranium is the same as that in the natural uranium, the value of the reprocessed uranium becomes lower than that of the natural uranium due to the produced U234 and U236. Accordingly, the value of the reprocessed uranium can be lower than that of the natural uranium depending on the composition of the reprocessed uranium.
[00681 Basically, in a nuclear reactor loading the natural uranium such as the heavy water reactor or the graphite reactor, when the infinite multiplication factor (criticality eigenvalue) of the reprocessed uranium is higher than that of the natural uranium throughout its lifetime, it can be said that the reprocessed uranium has the higher value. The infinite multiplication factor is the ratio between production reaction of the neutron and absorption reaction of the neutron in the reactor. The infinite 5 multiplication factor is 1 in the critical state, and the nuclear reactor requires the infinite multiplication factor of not less than 1. Because the ratio between neutron generation and neutron absorption of each uranium isotope in the fuel can be determined by (v x oa/of) using the number of neutrons generated by one 10 nuclear fission (v), the fission cross section (of), and the absorption cross section (oa) (the value is defined as a value coefficient), the sum total of products of the value coefficients of the individual uranium isotopes and the composition ratios of the individual uranium isotopes can be evaluated as the value of the 15 fuel.
[00691 Specifically, the value coefficients of the individual uranium isotopes contained in the reprocessed uranium, e.g., the value coefficients of the individual uranium isotopes in 0.0253 eV
Among the inventions according to the claims, it is possible to exert the effect of inventions according to any one of claims 6-10 most favorably.
[Effect of the Invention]
[0051]
In the present invention, since a spent fuel in a light water reactor is burned as the fuel for a heavy water reactor or a graphite reactor, it is possible to achieve a reduction in fuel cycle cost, an improvement in resource utilization rate, and a reduction in radioactive waste not only in a nuclear reactor using natural uranium or slightly enriched uranium but also in the light water reactor.
[Brief Description of the Drawings]
[0052]
FIG. 1 is a view conceptually showing an outer appearance of a fuel assembly for a CANDU reactor;
FIG. 2 is a view showing neutron flux spectrums (neutron energy distributions) in the CANDU reactor and a PWR;
FIG. 3 is a view showing changes in infinite multiplication factor in response to burnup of natural uranium and reprocessed uranium in the CANDU reactor;
FIG. 4 is a view conceptually showing changes in burnup and infinite multiplication factor when 6 types of reprocessed uranium reprocessed from fuels burned in the PWR and natural uranium are burned in the CANDU reactor; and FIG. 5 is a view conceptually showing changes in burnup and infinite multiplication factor when 6 types of reprocessed uranium reprocessed from fuels burned in a BWR and natural uranium are burned in the CANDU reactor.
[Explanations of Letters or Numerals]
[0053]
11 bearing pad 12 cladding tube 13 end support plate 14 fuel pellet spacer 16 end cap 15 [Best Mode for Carrying Out the Invention]
[0054]
A description will be given hereinbelow of embodiments of the present invention with reference to drawings. It is to be noted that the present invention is not limited to the following embodiments. Various modifications may be made to the following embodiments within the range identical or equivalent to the present invention.
[0055]
(With regard to use of reprocessed uranium in CANDU reactor) The present first embodiment relates to burning, in a CANDU reactor, of uranium reprocessed from a spent fuel in a PWR instead of natural uranium. A description will be given hereinbelow with reference to drawings and tables.
[0056]
In Table 1, there are shown a composition (%) of isotopes of uranium reprocessed from a fuel (uranium fuel) that has been discharged from a reactor as a spent fuel because the fuel has 5 attained a given burnup in a PWR having an electrical power of 1000 MWe, and has been left for 150 days, i.e., reprocessed uranium, together with a composition of isotopes of natural uranium.
[0057]
10 [Table 1]
uranium isotope natural uranium reprocessed uranium U234 0.0055 0.0121 U235 0.72 0.8259 U236 0 0.4379 U238 99.2745 98.7241 Total 100 100 [0058]
In Table 2, there are shown a fission cross section (of) and a absorption cross section (oa) of each of uranium isotopes in 15 thermal neutron energy (0.0253 eV) (Japanese Evaluated Nuclear Data Library JENDL-3.3, Japan Atomic Energy Agency, July 27, 2007). It is to be noted that the unit is barn.
[0059]
[Table 2]
uranium isotope fission cross section absorption cross section Of Ga U234 2.95x10-1 9.975x 101 U235 5.851x10-2 9.869x101 U236 6.129x 10.2 5.295 U238 1.177x105 2.717 [0060]
A fuel assembly for the CANDU reactor shown in FIG. 1 was designed using the natural uranium and the reprocessed uranium each having the composition shown in Table 1. In FIG. 1, reference numeral 11 denotes a bearing pad, reference numeral 12 denotes a cladding tube, reference numeral 13 denotes an end support plate, reference numeral 14 denotes a fuel pellet loaded in the cladding tube 12, reference numeral 15 denotes a spacer, and reference numeral 16 denotes an end cap. The end support plate 13 is fixed to an end portion of a fuel assembly 10 by resistance welding, the end cap 16 is fixed by resistance welding in order to seal an end portion of the cladding tube 12, and the spacer 15 and the bearing pad 11 are soldered to the cladding tube 12.
[0061]
The length of the fuel assembly 10 is 49.5 cm, the outer diameter of the cladding tube 12 is 13.061 mm, the wall thickness thereof is 0.419 mm, the diameter of the fuel pellet 15 is 12.154 mm, the length thereof is 16.40 mm, and the density thereof is 10.6 g/cm 3 .
[0062]
Next, for different fuels (the natural uranium and the reprocessed uranium) of the fuel assembly, neutron energy distributions (neutron flux spectrums) in the fuels were calculated. The result is shown in FIG. 2. In FIG. 2, the horizontal axis is a logarithmic scale of neutron energy (eV), while the vertical axis indicates a neutron density (Source/Lethargy/Volume) in the fuel. A thick line indicates the neutron flux spectrum of the natural uranium, while a thin line indicates that of the reprocessed uranium. The neutron flux spectrum in the PWR is also shown in FIG. 2 for reference purposes.
[0063]
In FIG. 2, the thick line can be barely distinguished from the thin line in a region from about 0.3 x 1E - 2 eV to about 0.3 X
1E - 1 eV, but they completely conform to each other in other regions. That is, the neutron spectrums of the natural uranium and the reprocessed uranium in the CANDU reactor are scarcely different from each other in all neutron energy regions, and it can be seen from the result that it is possible to burn the reprocessed uranium without being processed in the CANDU reactor. It is inferred that the reason why the neutron flux is larger in the CANDU reactor than in the PWR in the region where the energy is from 0.3 x 1E - 2 eV to 0.3 x 1E - 1 eV is that the CANDU reactor has a large reactor core so that the leakage of the neutron is small, and the moderating ratio of the heavy water as the moderator is large.
[00641 Subsequently, the calculation of burnup of the respective fuels was performed to compare changes in infinite multiplication factor in response to the burnup. The result is shown in FIG. 3.
In FIG. 3, the horizontal axis indicates the burnup (the burnup of the fuel assembly) (MWd/t), the vertical axis on the left indicates the infinite multiplication factor (k-infinity), and the vertical axis on the right indicates a difference between infinite multiplication factors of both fuels Ak (the infinite multiplication factor of the reprocessed uranium - that of the natural uranium) W. In the drawing, a line joining black diamonds indicates a natural uranium fuel, a line joining black squares indicates a reprocessed uranium fuel, and a line joining open circles indicates the difference between infinite multiplication factors of both fuels.
[00651 Although the currently used CANDU reactor adopts 7500 MWd/t as a discharge burnup, the discharge burnup of the reprocessed uranium corresponding to the infinite multiplication factor at this point is about 9500 MWd/t, as indicated by arrows in FIG. 3. This denotes that, when the reprocessed uranium fuel is used in the CANDU reactor, the fuel can be used for a period prolonged by about 25%, and it becomes possible to achieve a reduction in fuel cycle cost and a reduction in radioactive waste.
[00661 (With regard to value of reprocessed uranium) Next, a description will be given hereinbelow of a value coefficient of the uranium isotope and a value of the reprocessed uranium.
In Table 1, more U235 that is liable to undergo the nuclear fission reaction is contained in the reprocessed uranium than in the natural uranium and, as a result, it can be said that the reprocessed uranium is a fuel having a value higher than that of the natural uranium. However, the composition of the uranium isotopes in the uranium reprocessed from the light water reactor does not always conform to the composition in Table 1, and is dependent on the enrichment when loaded in the light water reactor, and the discharge burnup when discharged from the light water reactor.
[00671 As shown in Table 1, U234 of which only a minute amount is contained in the natural uranium and U236 that does not exist in the natural uranium are produced and contained in the reprocessed uranium, and they function as poisons for capturing neutrons. Consequently, for example, when the proportion of U235 in the reprocessed uranium is the same as that in the natural uranium, the value of the reprocessed uranium becomes lower than that of the natural uranium due to the produced U234 and U236. Accordingly, the value of the reprocessed uranium can be lower than that of the natural uranium depending on the composition of the reprocessed uranium.
[00681 Basically, in a nuclear reactor loading the natural uranium such as the heavy water reactor or the graphite reactor, when the infinite multiplication factor (criticality eigenvalue) of the reprocessed uranium is higher than that of the natural uranium throughout its lifetime, it can be said that the reprocessed uranium has the higher value. The infinite multiplication factor is the ratio between production reaction of the neutron and absorption reaction of the neutron in the reactor. The infinite 5 multiplication factor is 1 in the critical state, and the nuclear reactor requires the infinite multiplication factor of not less than 1. Because the ratio between neutron generation and neutron absorption of each uranium isotope in the fuel can be determined by (v x oa/of) using the number of neutrons generated by one 10 nuclear fission (v), the fission cross section (of), and the absorption cross section (oa) (the value is defined as a value coefficient), the sum total of products of the value coefficients of the individual uranium isotopes and the composition ratios of the individual uranium isotopes can be evaluated as the value of the 15 fuel.
[00691 Specifically, the value coefficients of the individual uranium isotopes contained in the reprocessed uranium, e.g., the value coefficients of the individual uranium isotopes in 0.0253 eV
20 as neutron usage energy in the case of the CANDU reactor, are determined in advance. Subsequently, on the basis of the products of the composition ratios of the individual uranium isotopes and the value coefficients of the individual uranium isotopes in the reprocessed uranium that are obtained by analysis, 25 the value of the reprocessed uranium when used as the fuel in the heavy water reactor or the like is calculated and compared with the value of the natural uranium that is prepared separately, and it is determined whether or not the reprocessed uranium is to be reused as the fuel in the heavy water reactor or the like. That is, when the value of the reprocessed uranium is higher than that of the natural uranium, it is determined that the reprocessed uranium has the value as the fuel for the heavy water reactor or the like and is reused. When the value thereof is lower, the reprocessed uranium is not used as the fuel for the heavy water reactor or the like, and is used for other uses, for example, enriching the reprocessed uranium again, using the reprocessed uranium for the production of a MOX fuel, and the like.
[00701 It is to be noted that the composition ratios of the individual uranium isotopes in the reprocessed uranium are determined by calculation or an experience using the composition and a burnup record of the spent fuel of the light water reactor to be used as the raw material for the reprocessed uranium. Values of the composition ratios to be adopted need to be further verified by measurement of radiations from the reprocessed uranium or the like.
[00711 Table 3 shows the value coefficient (vof/(Ya) for the thermal neutron (0.0253 eV) of each uranium isotope. It is to be noted that v in Table 3 indicates the average number of neutrons generated by one nuclear fission of each uranium isotope.
[00721 [Table 3]
uranium isotope value coefficient (vhf/va) U234 7.023x10-3 U235 2.085 U236 2.701x10-2 U238 1.058x10-5 [0073]
In Table 4, values of the reprocessed uranium and the natural uranium having the compositions shown in Table 1 as the fuels are shown for comparison purposes. As shown in Table 4, the value of the natural uranium is 1.50, while the value of the reprocessed uranium is 1.73. Therefore, it can be seen that the reprocessed uranium having the composition shown in Table 1 has the value higher than that of the natural uranium as the fuel, and is worth being burned in the CANDU reactor. Further, it also can be seen that there is little advantage in using the reprocessed uranium having the value of not more than 1.50 as the fuel for the CANDU reactor.
[0074]
[Table 4]
uranium isotope natural uranium reprocessed uranium U234 3.86x 10.5 8.50x 10 -5 U235 1.50 1.72 U236 0 1.18x10-2 U238 1.05x 10,3 1.04x 10,3 total (value) 1.50 1.73 [0075]
(Example 1) Next, as a specific example, several fuels using the reprocessed uranium having conditions different from those in Table 1 were evaluated using the value coefficient mentioned above and the like. The evaluation was performed for totally 12 types of reprocessed uranium including 6 types of reprocessed uranium obtained by discharging and recovering a STEP 1 fuel (the enrichment of 4.1%) and a STEP 2 fuel (the enrichment of 4.8%) at discharge burnups of 30 GWd/t, 40 GWd/t, and 50 GWd/t from the PWR, 3 types of reprocessed uranium obtained by discharging and recovering an 8 x 8 fuel (the enrichment of 3.6%) at discharge burnups of 30 GWd/t, 40 GWd/t, and 50 GWd/t from the BWR, and 3 types of reprocessed uranium obtained by discharging and recovering a 9 x 9 fuel (the enrichment of 4%) at discharge burnups of 35 GWd/t, 45 GWd/t, and 55 GWd/t also from the BWR. A fuel assembly for the CANDU reactor was produced using the 12 types of the reprocessed uranium and the natural uranium, and was operated.
[0076]
Table 5 shows values of the reprocessed uranium as the fuels. It is to be noted that values surrounded by ( ) in sections of the discharge burnup, e.g., (35) is the discharge burnup of the 9 x 9 fuel of the BWR.
[0077]
[Table 5]
type of fuel for light water discharge burnup[GWd/t]
reactor 30(35) 40(45) 50(55) 3.42 2.35 1.55 (enrichment of 4.1%) PWR
4.58 3.34 2.35 (enrichment of 4.8%) 8 x 8 fuel 2.58 1.60 0.92 BWR
9 x 9 fuel 2.56 1.56 0.87 [0078]
FIG. 4 conceptually shows (rough states of) the burnups and changes in infinite multiplication factor of totally 6 types of the reprocessed uranium reprocessed from the 6 types of fuels burned in the PWR and the natural uranium when they are burned in the CANDU reactor. FIG. 5 conceptually shows the burnups and changes in infinite multiplication factor of totally 6 types of uranium reprocessed from the 6 types of fuels burned in the BWR and the natural uranium when they are burned in the CANDU reactor. In each of FIGS. 4 and 5, the horizontal axis indicates the burnup (MWd/t), while the vertical axis indicates the infinite multiplication factor (k-infinity).
[00791 In FIG. 4, a line indicated by reference numeral 1 shows the case of the STEP 2 fuel at the discharge burnup of 30 GWd/t, a line indicated by reference numeral 2 shows the case of the STEP
5 2 fuel at the discharge burnup of 40 GWd/t and the case of the STEP 1 fuel at the discharge burnup of 30 GWd/t, a line indicated by reference numeral 3 shows the case of the STEP 2 fuel at the discharge burnup of 50 GWd/t and the case of the STEP 1 fuel at the discharge burnup of 40 GWd/t, and a line indicated by 10 reference numeral 4 shows the case of the STEP 1 fuel at the discharge burnup of 50 GWd/t and the case of the natural uranium.
[00801 In addition, in FIG. 5, a line indicated by reference numeral 15 1 shows the case of the 8 x 8 fuel at the discharge burnup of 30 GWd/t and the case of the 9 x 9 fuel at the discharge burnup of 35 GWd/t, a line indicated by reference numeral 2 shows the case of the 8 x 8 fuel at the discharge burnup of 40 GWd/t, the case of the 9 x 9 fuel at the discharge burnup of 45 GWd/t, and the case of the 20 natural uranium, and a line indicated by reference numeral 3 shows the case of the 8 x 8 fuel at the discharge burnup of 50 GWd/t and the case of the 9 x 9 fuel at the discharge burnup of 55 GWd/t.
[00811 25 It can be seen from Table 5 and FIG. 4 that, broadly speaking, the reprocessed uranium from the spent fuel of the PWR
has a value higher than that of the natural uranium. Further, it can be seen from Table 5 and FIG. 5 that the reprocessed uranium from the spent fuel of the BWR has a value equivalent to that of the natural uranium fuel when the fuel is discharged at the burnup of not more than about the limit of the attained burnup -GWd/t (40 GWd/t for the 8 x 8 fuel, 45 GWd/t for the 9 x 9 fuel).
5 [00821 (Example 2) The present example relates to the use of the spent fuel discharged from the light water reactor in the graphite reactor.
In a magnox reactor or an improved gas-cooled reactor, graphite 10 having the moderating ratio higher than that of the light water is used as the moderator, and hence the natural uranium fuel is used similarly to the heavy water reactor. Accordingly, it is possible to use the reprocessed uranium fuel in this case as well.
However, because of a difference between the moderating ratio of the heavy water and that of graphite, and a difference in the structure of the nuclear reactor resulting from a difference in physical condition, the graphite reactor is slightly different from the heavy water reactor in some cases in the neutron spectrum, the number of neutrons generated by one nuclear fission, and the fission cross section and the absorption cross section of the neutron from the generation to the disappearance. As a result, the value used for the evaluation of the reprocessed uranium fuel may be slightly different from the value targeted for the heavy water reactor.
[00831 However, basics are the same as those of the CANDU
reactor, and all data items required for study of the case where the reprocessed uranium is burned in place of the natural uranium fuel in the graphite reactor are disclosed in, e.g., Japanese Evaluated Nuclear Data Library JENDL-3.3 or the like of Japan Atomic Energy Agency mentioned above. In addition, various programs for various graphite reactors such as a program for calculating the change in infinite multiplication factor of the fuel in response to progress in burnup of the natural uranium fuel and the like are basically identical with those for light water reactors, and the programs are already developed and used.
Alternatively, although values and data items to be inputted are different, it is possible to use the programs for light water reactors. Accordingly, it is possible to perform required calculation by inputting values of the reprocessed uranium fuel into the programs instead of values (data items) of the natural uranium fuel. Consequently, it is omitted to show specific values particularly for a specific graphite moderated reactor as an example and give a description using drawings and tables.
[0084]
(Example 3) The present example relates to reburning of plutonium and uranium reprocessed from the spent fuel in the light water in a heavy water reactor or the like without separating plutonium from uranium.
[0085]
The description up to example 2 has related to the reburning of only uranium reprocessed from the spent fuel in the light water reactor in a heavy water reactor or the like, and the fuel mentioned above contains about 1% of plutonium (mainly Pu239) as the fissile material. In this example, daughter = CA 02724582 2010-11-16 nuclides generated by the nuclear fission are removed from the spent fuel in the light water reactor, e.g., the BWR, and uranium and plutonium that are left (reprocessed) are used as the fuel for the heavy water reactor without separating plutonium from uranium. Consequently, the fuel of the present example has a value higher than that of the fuel using only reprocessed uranium, and it is possible to use the fuel in the type of a reactor in which slightly enriched uranium is used with no problem.
[00861 However, when compared with Example 1, plutonium is different from uranium in nuclear characteristics such as the number of neutrons generated by the nuclear fission (plutonium generates more neutrons than uranium), the fission cross section, and the absorption cross section of the neutron, and is also different from uranium in the neutron spectrum in the reactor.
[00871 However, basics are the same as those in Example 1. All data items required for the study of the case where a reprocessed uranium- plutonium composite fuel is burned in place of the reprocessed uranium fuel, specifically nuclear characteristics of nuclides of plutonium isotopes, are disclosed in many laboratories and books, e.g., Japanese Evaluated Nuclear Data Library JENDL-3.3 or the like of Japan Atomic Energy Agency mentioned above because the nuclear characteristics are necessary for the development of fast breeder reactors. Further, various programs for the heavy water reactor and the graphite reactor such as a program for calculating the change in the infinite multiplication factor of the fuel in response to progress in burnup of the reprocessed uranium-plutonium fuel and the like are already developed. That is, although information and values to be inputted such as the shape of the reactor core and the like are slightly different, basic calculation processes such as dividing the reactor core into meshes and determining the nuclear reaction and the neutron flux in unit time in each mesh by calculation, exchanging calculated values among the individual meshes, and repeating the same calculation processing using values modified by performing the exchange in the next unit time are the same as those in the case of the light water reactor. Therefore it is possible to easily perform the calculation required for the present example.
[0088]
Further, in the BWR, plutonium is actually burned together with uranium. That is, even in the currently used BWR in which the uranium fuel is burned, at the end of each operation cycle, the proportion of nuclear fission of Pu239 generated by capturing of neutrons by U238 is larger than that of U235 in the actual nuclear fission. Consequently, there is no particular safety problem in burning the reprocessed uranium-plutonium fuel in the CANDU
reactor or the like. Consequently, similarly to Example 3, it is omitted to show specific values for the CANDU reactor or the reactor in which slightly enriched uranium is burned, e.g., a certain type of the graphite reactor as an example, and give a description using drawings and tables.
[0089]
As described in Example 1, the reactor core in the BWR is larger than that in the PWR, and flexibility in disposing the fuel in the reactor at the beginning of each cycle operation is high.
Therefore, the fuel discharged from the BWR as the spent fuel has the attained burnup close to the maximum value thereof (the permissible maximum burnup) in many cases. As a result, the 5 fuel using only uranium reprocessed from the spent fuel of the BWR has a value lower than that of the fuel using the natural uranium in many cases. However, the fuel using the reprocessed uranium-plutonium creates a value higher than that of the natural uranium, and can be sufficiently used not only in the 10 CANDU reactor in which the natural uranium is burned but also in a reactor using slightly enriched uranium having 1 to 2% of U235.
[00901 (Example 4) 15 The present example is a modification of Example 3 described above, and relates to blending of plutonium or uranium discharged from a disassembled nuclear bomb into the reprocessed uranium-plutonium and burning of the reprocessed uranium -plutonium.
20 [00911 As described above, the value of the reprocessed uranium-plutonium from the BWR is lower than that of the reprocessed uranium-plutonium from the PWR in general.
However, Pu239 and U235 used in the nuclear bomb have a 25 concentration of not less than 90%. Consequently, when a small amount of Pu239 or U235 discharged from a discarded nuclear bomb is blended into the reprocessed uranium-plutonium from the BWR, the value of the fuel is significantly increased, and the fuel becomes a fuel having the optimum value in the case where the fuel is burned in a heavy water reactor or a graphite reactor. In addition, it is possible to achieve effective utilization and peaceful disposal of the nuclear bomb.
[0092]
(Example 5) The present example is also a modification of Example 3 described above, and relates to burning of the reprocessed uranium-plutonium from the BWR in the CANDU reactor or the like with attention focused only on plutonium. That is, unlike U235 in the reprocessed uranium, the concentration of Pu239 in the reprocessed plutonium, i.e., the concentration of the plutonium isotope that actually undergoes nuclear fission is inherently high. Accordingly, there is no need to enrich Pu239 in the reprocessed plutonium. Consequently, when the reprocessed plutonium is burned in, e.g., the CANDU reactor, by blending the reprocessed plutonium into the natural uranium or the reprocessed uranium-plutonium and burning the reprocessed plutonium, it is possible to achieve a further improvement in discharge burnup.
[0093]
In this case, the reprocessed uranium is left over. The surplus reprocessed uranium is subjected to processing such as loading the surplus uranium in the fast breeder reactor, temporarily storing the surplus uranium in a storage facility, and the like.
[00701 It is to be noted that the composition ratios of the individual uranium isotopes in the reprocessed uranium are determined by calculation or an experience using the composition and a burnup record of the spent fuel of the light water reactor to be used as the raw material for the reprocessed uranium. Values of the composition ratios to be adopted need to be further verified by measurement of radiations from the reprocessed uranium or the like.
[00711 Table 3 shows the value coefficient (vof/(Ya) for the thermal neutron (0.0253 eV) of each uranium isotope. It is to be noted that v in Table 3 indicates the average number of neutrons generated by one nuclear fission of each uranium isotope.
[00721 [Table 3]
uranium isotope value coefficient (vhf/va) U234 7.023x10-3 U235 2.085 U236 2.701x10-2 U238 1.058x10-5 [0073]
In Table 4, values of the reprocessed uranium and the natural uranium having the compositions shown in Table 1 as the fuels are shown for comparison purposes. As shown in Table 4, the value of the natural uranium is 1.50, while the value of the reprocessed uranium is 1.73. Therefore, it can be seen that the reprocessed uranium having the composition shown in Table 1 has the value higher than that of the natural uranium as the fuel, and is worth being burned in the CANDU reactor. Further, it also can be seen that there is little advantage in using the reprocessed uranium having the value of not more than 1.50 as the fuel for the CANDU reactor.
[0074]
[Table 4]
uranium isotope natural uranium reprocessed uranium U234 3.86x 10.5 8.50x 10 -5 U235 1.50 1.72 U236 0 1.18x10-2 U238 1.05x 10,3 1.04x 10,3 total (value) 1.50 1.73 [0075]
(Example 1) Next, as a specific example, several fuels using the reprocessed uranium having conditions different from those in Table 1 were evaluated using the value coefficient mentioned above and the like. The evaluation was performed for totally 12 types of reprocessed uranium including 6 types of reprocessed uranium obtained by discharging and recovering a STEP 1 fuel (the enrichment of 4.1%) and a STEP 2 fuel (the enrichment of 4.8%) at discharge burnups of 30 GWd/t, 40 GWd/t, and 50 GWd/t from the PWR, 3 types of reprocessed uranium obtained by discharging and recovering an 8 x 8 fuel (the enrichment of 3.6%) at discharge burnups of 30 GWd/t, 40 GWd/t, and 50 GWd/t from the BWR, and 3 types of reprocessed uranium obtained by discharging and recovering a 9 x 9 fuel (the enrichment of 4%) at discharge burnups of 35 GWd/t, 45 GWd/t, and 55 GWd/t also from the BWR. A fuel assembly for the CANDU reactor was produced using the 12 types of the reprocessed uranium and the natural uranium, and was operated.
[0076]
Table 5 shows values of the reprocessed uranium as the fuels. It is to be noted that values surrounded by ( ) in sections of the discharge burnup, e.g., (35) is the discharge burnup of the 9 x 9 fuel of the BWR.
[0077]
[Table 5]
type of fuel for light water discharge burnup[GWd/t]
reactor 30(35) 40(45) 50(55) 3.42 2.35 1.55 (enrichment of 4.1%) PWR
4.58 3.34 2.35 (enrichment of 4.8%) 8 x 8 fuel 2.58 1.60 0.92 BWR
9 x 9 fuel 2.56 1.56 0.87 [0078]
FIG. 4 conceptually shows (rough states of) the burnups and changes in infinite multiplication factor of totally 6 types of the reprocessed uranium reprocessed from the 6 types of fuels burned in the PWR and the natural uranium when they are burned in the CANDU reactor. FIG. 5 conceptually shows the burnups and changes in infinite multiplication factor of totally 6 types of uranium reprocessed from the 6 types of fuels burned in the BWR and the natural uranium when they are burned in the CANDU reactor. In each of FIGS. 4 and 5, the horizontal axis indicates the burnup (MWd/t), while the vertical axis indicates the infinite multiplication factor (k-infinity).
[00791 In FIG. 4, a line indicated by reference numeral 1 shows the case of the STEP 2 fuel at the discharge burnup of 30 GWd/t, a line indicated by reference numeral 2 shows the case of the STEP
5 2 fuel at the discharge burnup of 40 GWd/t and the case of the STEP 1 fuel at the discharge burnup of 30 GWd/t, a line indicated by reference numeral 3 shows the case of the STEP 2 fuel at the discharge burnup of 50 GWd/t and the case of the STEP 1 fuel at the discharge burnup of 40 GWd/t, and a line indicated by 10 reference numeral 4 shows the case of the STEP 1 fuel at the discharge burnup of 50 GWd/t and the case of the natural uranium.
[00801 In addition, in FIG. 5, a line indicated by reference numeral 15 1 shows the case of the 8 x 8 fuel at the discharge burnup of 30 GWd/t and the case of the 9 x 9 fuel at the discharge burnup of 35 GWd/t, a line indicated by reference numeral 2 shows the case of the 8 x 8 fuel at the discharge burnup of 40 GWd/t, the case of the 9 x 9 fuel at the discharge burnup of 45 GWd/t, and the case of the 20 natural uranium, and a line indicated by reference numeral 3 shows the case of the 8 x 8 fuel at the discharge burnup of 50 GWd/t and the case of the 9 x 9 fuel at the discharge burnup of 55 GWd/t.
[00811 25 It can be seen from Table 5 and FIG. 4 that, broadly speaking, the reprocessed uranium from the spent fuel of the PWR
has a value higher than that of the natural uranium. Further, it can be seen from Table 5 and FIG. 5 that the reprocessed uranium from the spent fuel of the BWR has a value equivalent to that of the natural uranium fuel when the fuel is discharged at the burnup of not more than about the limit of the attained burnup -GWd/t (40 GWd/t for the 8 x 8 fuel, 45 GWd/t for the 9 x 9 fuel).
5 [00821 (Example 2) The present example relates to the use of the spent fuel discharged from the light water reactor in the graphite reactor.
In a magnox reactor or an improved gas-cooled reactor, graphite 10 having the moderating ratio higher than that of the light water is used as the moderator, and hence the natural uranium fuel is used similarly to the heavy water reactor. Accordingly, it is possible to use the reprocessed uranium fuel in this case as well.
However, because of a difference between the moderating ratio of the heavy water and that of graphite, and a difference in the structure of the nuclear reactor resulting from a difference in physical condition, the graphite reactor is slightly different from the heavy water reactor in some cases in the neutron spectrum, the number of neutrons generated by one nuclear fission, and the fission cross section and the absorption cross section of the neutron from the generation to the disappearance. As a result, the value used for the evaluation of the reprocessed uranium fuel may be slightly different from the value targeted for the heavy water reactor.
[00831 However, basics are the same as those of the CANDU
reactor, and all data items required for study of the case where the reprocessed uranium is burned in place of the natural uranium fuel in the graphite reactor are disclosed in, e.g., Japanese Evaluated Nuclear Data Library JENDL-3.3 or the like of Japan Atomic Energy Agency mentioned above. In addition, various programs for various graphite reactors such as a program for calculating the change in infinite multiplication factor of the fuel in response to progress in burnup of the natural uranium fuel and the like are basically identical with those for light water reactors, and the programs are already developed and used.
Alternatively, although values and data items to be inputted are different, it is possible to use the programs for light water reactors. Accordingly, it is possible to perform required calculation by inputting values of the reprocessed uranium fuel into the programs instead of values (data items) of the natural uranium fuel. Consequently, it is omitted to show specific values particularly for a specific graphite moderated reactor as an example and give a description using drawings and tables.
[0084]
(Example 3) The present example relates to reburning of plutonium and uranium reprocessed from the spent fuel in the light water in a heavy water reactor or the like without separating plutonium from uranium.
[0085]
The description up to example 2 has related to the reburning of only uranium reprocessed from the spent fuel in the light water reactor in a heavy water reactor or the like, and the fuel mentioned above contains about 1% of plutonium (mainly Pu239) as the fissile material. In this example, daughter = CA 02724582 2010-11-16 nuclides generated by the nuclear fission are removed from the spent fuel in the light water reactor, e.g., the BWR, and uranium and plutonium that are left (reprocessed) are used as the fuel for the heavy water reactor without separating plutonium from uranium. Consequently, the fuel of the present example has a value higher than that of the fuel using only reprocessed uranium, and it is possible to use the fuel in the type of a reactor in which slightly enriched uranium is used with no problem.
[00861 However, when compared with Example 1, plutonium is different from uranium in nuclear characteristics such as the number of neutrons generated by the nuclear fission (plutonium generates more neutrons than uranium), the fission cross section, and the absorption cross section of the neutron, and is also different from uranium in the neutron spectrum in the reactor.
[00871 However, basics are the same as those in Example 1. All data items required for the study of the case where a reprocessed uranium- plutonium composite fuel is burned in place of the reprocessed uranium fuel, specifically nuclear characteristics of nuclides of plutonium isotopes, are disclosed in many laboratories and books, e.g., Japanese Evaluated Nuclear Data Library JENDL-3.3 or the like of Japan Atomic Energy Agency mentioned above because the nuclear characteristics are necessary for the development of fast breeder reactors. Further, various programs for the heavy water reactor and the graphite reactor such as a program for calculating the change in the infinite multiplication factor of the fuel in response to progress in burnup of the reprocessed uranium-plutonium fuel and the like are already developed. That is, although information and values to be inputted such as the shape of the reactor core and the like are slightly different, basic calculation processes such as dividing the reactor core into meshes and determining the nuclear reaction and the neutron flux in unit time in each mesh by calculation, exchanging calculated values among the individual meshes, and repeating the same calculation processing using values modified by performing the exchange in the next unit time are the same as those in the case of the light water reactor. Therefore it is possible to easily perform the calculation required for the present example.
[0088]
Further, in the BWR, plutonium is actually burned together with uranium. That is, even in the currently used BWR in which the uranium fuel is burned, at the end of each operation cycle, the proportion of nuclear fission of Pu239 generated by capturing of neutrons by U238 is larger than that of U235 in the actual nuclear fission. Consequently, there is no particular safety problem in burning the reprocessed uranium-plutonium fuel in the CANDU
reactor or the like. Consequently, similarly to Example 3, it is omitted to show specific values for the CANDU reactor or the reactor in which slightly enriched uranium is burned, e.g., a certain type of the graphite reactor as an example, and give a description using drawings and tables.
[0089]
As described in Example 1, the reactor core in the BWR is larger than that in the PWR, and flexibility in disposing the fuel in the reactor at the beginning of each cycle operation is high.
Therefore, the fuel discharged from the BWR as the spent fuel has the attained burnup close to the maximum value thereof (the permissible maximum burnup) in many cases. As a result, the 5 fuel using only uranium reprocessed from the spent fuel of the BWR has a value lower than that of the fuel using the natural uranium in many cases. However, the fuel using the reprocessed uranium-plutonium creates a value higher than that of the natural uranium, and can be sufficiently used not only in the 10 CANDU reactor in which the natural uranium is burned but also in a reactor using slightly enriched uranium having 1 to 2% of U235.
[00901 (Example 4) 15 The present example is a modification of Example 3 described above, and relates to blending of plutonium or uranium discharged from a disassembled nuclear bomb into the reprocessed uranium-plutonium and burning of the reprocessed uranium -plutonium.
20 [00911 As described above, the value of the reprocessed uranium-plutonium from the BWR is lower than that of the reprocessed uranium-plutonium from the PWR in general.
However, Pu239 and U235 used in the nuclear bomb have a 25 concentration of not less than 90%. Consequently, when a small amount of Pu239 or U235 discharged from a discarded nuclear bomb is blended into the reprocessed uranium-plutonium from the BWR, the value of the fuel is significantly increased, and the fuel becomes a fuel having the optimum value in the case where the fuel is burned in a heavy water reactor or a graphite reactor. In addition, it is possible to achieve effective utilization and peaceful disposal of the nuclear bomb.
[0092]
(Example 5) The present example is also a modification of Example 3 described above, and relates to burning of the reprocessed uranium-plutonium from the BWR in the CANDU reactor or the like with attention focused only on plutonium. That is, unlike U235 in the reprocessed uranium, the concentration of Pu239 in the reprocessed plutonium, i.e., the concentration of the plutonium isotope that actually undergoes nuclear fission is inherently high. Accordingly, there is no need to enrich Pu239 in the reprocessed plutonium. Consequently, when the reprocessed plutonium is burned in, e.g., the CANDU reactor, by blending the reprocessed plutonium into the natural uranium or the reprocessed uranium-plutonium and burning the reprocessed plutonium, it is possible to achieve a further improvement in discharge burnup.
[0093]
In this case, the reprocessed uranium is left over. The surplus reprocessed uranium is subjected to processing such as loading the surplus uranium in the fast breeder reactor, temporarily storing the surplus uranium in a storage facility, and the like.
Claims (12)
1. A nuclear reactor fuel for use in a heavy water reactor or a graphite reactor that is produced by using a fuel material reprocessed from a spent fuel in a light water reactor without enriching U235 contained in the fuel material.
2. The nuclear reactor fuel according to claim 1, wherein the fuel material reprocessed from the spent fuel in the light water reactor is uranium.
3. The nuclear reactor fuel according to claim 2, wherein the light water reactor is a PWR.
4. The nuclear reactor fuel according to claim 1 or claim 2, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
5. The nuclear reactor fuel according to claim 3, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
6. A process for producing a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor comprising the fuel material recovering step of recovering a fuel material from a spent fuel in a light water reactor, and the fuel producing step of producing the fuel by using the reprocessed fuel material without enriching U235 contained in the fuel material.
7. The process for producing a nuclear reactor fuel according to claim 6 further comprising, after the fuel material recovering step, the composition ratio calculating step of calculating a composition ratio of each composition material in the reprocessed fuel material, the value coefficient calculating step of calculating a quotient by dividing a product of a fission cross section and a number of neutrons generated by nuclear fission by captured cross section for each composition material to determine the quotient as a value coefficient for each composition material, the value calculating step of calculating a product of the composition ratio and the value coefficient for each composition material to determine a sum total of the product as a value of the reprocessed fuel material, and the reuse determining step of comparing the value of the fuel material with a given reference value to determine whether or not the fuel producing step is to be performed according to a magnitude relationship therebetween.
8. A process for producing a nuclear reactor fuel for use in a heavy water reactor or a graphite reactor comprising the uranium recovering step of recovering uranium from a spent fuel in a light water reactor, and the fuel producing step of producing the fuel by using the reprocessed uranium without enriching U235 contained in the reprocessed uranium as a composition material.
9. The process for producing a nuclear reactor fuel of the invention according to claim 8 comprising, after the uranium recovering step, the composition ratio calculating step of calculating a composition ratio of each uranium isotope in the reprocessed uranium, the value coefficient calculating step of calculating a quotient by dividing a product of a fission cross section and a number of neutrons generated by nuclear fission by a absorption cross section for each uranium isotope to determine the quotient as a value coefficient of each uranium isotope, the value calculating step of calculating a product of the composition ratio and the value coefficient for each uranium isotope to determine a sum total of the product as a value of the reprocessed uranium, and the reuse determining step of comparing the value of the reprocessed uranium with a given reference value to determine whether or not the fuel producing step is to be performed according to a magnitude relationship therebetween.
10. The process for producing a nuclear reactor fuel according to any one of claims 6-9, wherein the light water reactor is a PWR.
11. The process for producing a nuclear reactor fuel according to any one of claims 6-9, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
12. The process for producing a nuclear reactor fuel according to claim 10, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| PCT/JP2008/060569 WO2009150710A1 (en) | 2008-06-09 | 2008-06-09 | Fuel for heavy-water reactor or graphite reactor and process for producing the same |
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| CN108962417B (en) * | 2018-06-22 | 2020-02-21 | 中核核电运行管理有限公司 | Heavy water reactor cobalt isotope production method |
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| JP3008129B2 (en) * | 1990-11-29 | 2000-02-14 | 株式会社日立製作所 | Manufacturing method of fuel assembly |
| US5475723A (en) * | 1994-03-21 | 1995-12-12 | General Electric Company | Nuclear fuel cladding with hydrogen absorbing inner liner |
| TW258815B (en) * | 1994-07-13 | 1995-10-01 | Gen Electric | Nuclear fuel cladding having a gold coating |
| JP3325462B2 (en) * | 1996-06-04 | 2002-09-17 | 核燃料サイクル開発機構 | Manufacturing method of fuel assembly |
| JP2000292575A (en) * | 1999-04-08 | 2000-10-20 | Hitachi Ltd | Fuel assembly |
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2008
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| KR20110070962A (en) | 2011-06-27 |
| KR101488555B1 (en) | 2015-02-02 |
| WO2009150710A1 (en) | 2009-12-17 |
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