WO2009150710A1 - Combustible pour réacteur à eau lourde ou réacteur à graphite et son procédé de production - Google Patents

Combustible pour réacteur à eau lourde ou réacteur à graphite et son procédé de production Download PDF

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Publication number
WO2009150710A1
WO2009150710A1 PCT/JP2008/060569 JP2008060569W WO2009150710A1 WO 2009150710 A1 WO2009150710 A1 WO 2009150710A1 JP 2008060569 W JP2008060569 W JP 2008060569W WO 2009150710 A1 WO2009150710 A1 WO 2009150710A1
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Prior art keywords
fuel
uranium
reactor
recovered
value
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PCT/JP2008/060569
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English (en)
Japanese (ja)
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靖典 大岡
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原子燃料工業株式会社
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Priority to PCT/JP2008/060569 priority Critical patent/WO2009150710A1/fr
Priority to CA2724582A priority patent/CA2724582A1/fr
Priority to KR1020107028123A priority patent/KR101488555B1/ko
Publication of WO2009150710A1 publication Critical patent/WO2009150710A1/fr

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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
    • 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

Definitions

  • the present invention relates to a fuel for a heavy water reactor or a graphite reactor and a method for producing the same, and more particularly to a fuel material recovered from a spent fuel in a light water reactor without concentrating U235 contained in the fuel material.
  • the present invention relates to a fuel for a heavy water reactor or a graphite reactor manufactured using the same and a method for manufacturing the same.
  • a fuel in which U (uranium) 235 is concentrated to about 3.6 to 5% by mass (hereinafter simply referred to as “%”) is used as the fuel.
  • enriched uranium in which U (uranium) 235 is concentrated to about 3.6 to 5% by mass
  • % in which U (uranium) 235 is concentrated to about 3.6 to 5% by mass
  • the fuel is taken out of the furnace as spent fuel when it reaches a predetermined burnup, stored for a certain period, and then subjected to reprocessing.
  • a predetermined burnup in PWR, the upper limit (allowable burnup) of the ultimate burnup of the fuel with the enrichment of 4.1% and the fuel of 4.8% is 48 GWd / t and 55 GWd / t, respectively.
  • the burnup differs depending on the fuel loading position in the furnace, and since it cannot be taken out of the furnace while the operation continues, it is actually taken out in a state of about 40 GWd / t and 45 GWd / t, respectively. Yes.
  • BWR the upper limit of the ultimate burn-up of fuel with enrichment of about 3.6% and fuel of about 4.0% is 48 GWd / t and 55 GWd / t, respectively, and BWR has a larger core than PWR. Therefore, since the degree of freedom of arrangement when loading each fuel in the furnace is high, all the fuels are taken out with the ultimate burnup close to these.
  • the enrichment value is set to the above value because if it is too high, the cost of enrichment is too high, and the reactivity of the core immediately after loading the new fuel becomes too high (the infinite increase described later). This is in consideration of the fact that the magnification is too large).
  • the upper limit of the ultimate burnup is that U235 decreases with combustion, and the number of nuclides that absorb neutrons generated by fission increases, making it difficult to keep the core critical, fuel pellets and fuel rods This is because it is necessary to prevent damage to the cladding tube.
  • the only commercial reactor is a light water reactor, but in the world, mainly in Canada, it uses natural uranium containing only 0.72% of U235 and slightly enriched uranium enriched to 1% to 2% as fuel.
  • a reactor using heavy water as a moderator for example, a CANDU furnace
  • other types of reactors using graphite as a moderator are operating as commercial reactors.
  • the reason why such a nuclear reactor is used is that the reduction ratio of heavy water (average reduction rate of logarithm of energy per collision ⁇ scattering cross section / absorption cross section) is 80 times that of light water, and that of graphite is 2.4. For this reason, in a CANDU furnace using heavy water, the core can be kept critical even with natural uranium, and it is not necessary to use expensive enriched uranium as fuel.
  • the fuel removal burn-up is much smaller than that of a light water reactor. For example, it is about 7.5 GWd / t in a CANDU furnace in which a total of 32 units are operating in seven countries such as Canada and Korea.
  • the fuel taken out of the reactor as spent fuel because it has reached a predetermined burnup is a radioactive element (daughter) generated by U235 fission. Because strong radiation is emitted from the nuclide) and there is also heat generation, it is stored for a certain period in a storage pool with the prescribed equipment, and then recovers unburned uranium and plutonium generated by uranium combustion Reprocessing such as performing (Non-Patent Document 1). Seihei Kiyose, “Chemical Industry of Spent Fuel and Plutonium”, published by Nikkan Kogyo Shimbun, 1984
  • the present invention has been made for the purpose of solving the above-described problems, and fuel used in a light water reactor is burned in a nuclear reactor using natural uranium or micro-enriched uranium.
  • the invention of each claim will be described below.
  • the invention described in claim 1 A nuclear fuel used in heavy water reactors or graphite reactors, A fuel for a nuclear reactor, which is manufactured by using a fuel substance recovered from fuel used in a light water reactor without concentrating U235 contained in the fuel substance. .
  • the fuel substance recovered from the fuel used in the light water reactor is originally natural uranium or slightly enriched uranium without concentrating U235 contained in the fuel substance. Is burned again in a heavy water reactor or graphite furnace using as a fuel, the fuel cycle cost can be reduced, the resource utilization rate can be improved, and the radioactive waste can be reduced in the heavy water reactor or graphite furnace.
  • light water reactors can also reduce enrichment costs, reduce fuel cycle costs by eliminating the need for enrichment processes, effectively use resources, and reduce radioactive waste.
  • the “spent fuel” means not only the fuel that has reached the upper limit of the burnup degree or a burnup close to it, but also the next cycle run, even if you try to burn again in the next cycle run. Since the allowable maximum burnup is reached halfway, the fuel that cannot be burned in the next cycle operation is included.
  • uranium fuel enriched with natural uranium is mainly used in current light water reactors, but it is planned to use MOX fuel blended with plutonium in the future. In the future, thorium fuel is also under development. For this reason, it is not limited to uranium fuel.
  • Fluel material refers to fissile material and its isotopes. In our current light water reactor, U235 and Pu239 and their isotopes are excluded in the future. In addition, other impurity elements that may be inevitably contained may be included.
  • “manufactured using the U235 contained in the fuel substance without concentrating” means that it has been used unless the U235 in uranium recovered from the used fuel is concentrated.
  • “manufacturing using U235 without concentrating” means not concentrating U235 in recovered uranium when manufacturing fuel for heavy water reactors or graphite reactors (unlike natural uranium, recovered uranium ) Contains U232, which is a source of intense radioactivity, complicating the concentration process and increasing costs).
  • the fuel pellets in the fuel rods constituting the nuclear fuel assembly also correspond to the “reactor fuel” of the present invention, and even if there is only one fuel pellet, the “fuel for the nuclear reactor” of the present invention. Applicable.
  • the “heavy water reactor” and “graphite furnace” include a nuclear reactor that uses gas or light water as a coolant as long as heavy water or graphite is used as a moderator.
  • the nuclear reactor fuel for the heavy water reactor or graphite reactor is manufactured using uranium recovered from the fuel that has been used in the light water reactor, the U235 remaining unburned in the light water reactor is removed. Mainly, other trace amounts of uranium isotopes are burned in the heavy water reactor or graphite furnace, and the fuel cost is reduced not only for the heavy water reactor or graphite furnace but also for the light water reactor.
  • the fuel used in the light water reactor contains a larger amount of fissile material U235 than natural uranium, the effect of the invention of claim 1 can be exhibited particularly well.
  • the invention described in claim 3 3.
  • the effect of the invention of claim 1 or claim 2 can be exhibited more satisfactorily.
  • the PWR has a smaller core than the BWR, and further has a low degree of freedom in the arrangement of fuel in the reactor. Therefore, as described above, the allowable maximum burnup must be used with a certain margin. A lot of unobtainable fuel is generated. As a result, the composition ratio of U235 in the spent fuel is higher than that in natural uranium. When used in heavy water reactors and graphite furnaces, the burnup can be increased and the fuel exchange interval is lengthened. Because it can.
  • the invention according to claim 4 The nuclear reactor fuel according to claim 1 or 2, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
  • the invention according to claim 5 The reactor fuel according to claim 3, wherein the reactor fuel is a fuel for a CANDU reactor.
  • the invention described in claim 6 A method for producing a fuel for a nuclear reactor used in a heavy water reactor or a graphite reactor, A fuel material recovery step for recovering the fuel material from the fuel used in the light water reactor; And a fuel production step for producing a fuel by using the recovered fuel substance without concentrating U235 contained in the fuel substance. Is the method.
  • the invention of this claim captures the invention of claim 1 as a method invention.
  • a composition ratio calculating step for obtaining a composition ratio of each composition substance in the recovered fuel substance; For each composition material, a product of the fission cross section and the number of neutrons generated by the fission is divided by the capture cross section, and a value coefficient calculation step for setting the value coefficient of each composition material with its quotient; For each of the composition materials, a product of the composition ratio and the value coefficient is obtained, and a value calculation step for obtaining the value of the recovered fuel material with the sum of the product, and 7.
  • a reuse determination step of comparing the value of the fuel substance with a predetermined reference value and determining the shift to the fuel production step according to the magnitude of the value.
  • the value of the recovered fuel substance is compared with a predetermined reference value, it is determined whether or not the fuel substance is used as a fuel for a heavy water reactor or a light water reactor.
  • the burn-up degree is increased, and the fuel exchange interval can be extended.
  • the fission cross section ( ⁇ f ), the number of neutrons generated by fission ( ⁇ ), and the capture cross section ( ⁇ a ) depend on the neutron energy.
  • composition ratio of each composition substance in the recovered fuel substance may be the composition ratio of the uranium isotope and the plutonium isotope in the mixture of uranium and plutonium.
  • calculation of the composition ratio any means such as calculation, actual measurement, and experience may be used.
  • the “predetermined reference value” is a composition ratio of each compositional substance targeting a fuelous substance that is originally burned in a nuclear reactor in which the fuelous substance is burned, for example, natural uranium in a CANDU furnace.
  • the fission cross section, the number of neutrons generated by fission, and the capture cross section, the value obtained by the same procedure as the fuel substance value calculation step in the invention of this claim may be used.
  • it may be a corrected value reflecting a difference in the neutron spectrum and the like, an error such as a measurement of the composition ratio, for example, a cost for using the fuel for the CANDU reactor.
  • the predetermined reference value is prepared by taking that amount into account.
  • the value of the fuel substance consisting of “recovered fuel substance” and “plutonium or U235 extracted from the nuclear bomb and blended” is calculated and compared with a predetermined reference value.
  • Good this is an invention equivalent to the present invention, and such a case is also included in the invention of this claim.
  • a plurality of reference values for heavy water reactors and graphite furnaces may be created, and the reactor to be reused may be changed according to the values.
  • the invention according to claim 8 provides: A method for producing a fuel for a nuclear reactor used in a heavy water reactor or a graphite reactor, A uranium recovery step for recovering uranium from fuel used in light water reactors; And a fuel production step for producing a fuel using the recovered uranium without concentrating U235 contained as a composition thereof. .
  • the invention of this claim captures the invention of claim 2 as a method invention.
  • the recycling determination step of comparing the value of the recovered uranium with a predetermined reference value, and determining the shift to the fuel production step according to the magnitude thereof. It is a manufacturing method of the fuel for nuclear reactors of description.
  • the value of the recovered uranium is compared with a predetermined reference value to determine whether or not the fuel substance is used as a heavy water reactor or light water reactor fuel.
  • uranium includes the case where other elements are unavoidably included. For this reason, when determining the value of recovered uranium, the influence of such an impurity element may be corrected based on experience or actual measurement.
  • the invention according to claim 10 is: The method for producing a fuel for a nuclear reactor according to any one of claims 6 to 9, wherein the light water reactor is a PWR.
  • the invention according to claim 11 The method for producing a nuclear fuel according to any one of claims 6 to 9, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
  • the invention according to claim 12 is The method for producing a nuclear fuel according to claim 10, wherein the heavy water reactor or the graphite reactor is a CANDU reactor.
  • fuel used in light water reactors is burned as fuel for heavy water reactors or graphite reactors. Therefore, in light water reactors as well as nuclear reactors using natural uranium and microenriched uranium, fuel cycle costs are reduced and resource utilization is reduced. Improvement and reduction of radioactive waste.
  • the first embodiment relates to burning uranium recovered from fuel used in PWR in place of natural uranium in a CANDU furnace.
  • description will be made with reference to the drawings.
  • Table 1 shows that uranium recovered from fuel (uranium fuel) that has been taken out of the furnace as spent fuel and has passed 150 days since it has reached a predetermined burnup at a PWR of 1000 MWe, that is, isotope of recovered uranium.
  • the body composition (%) is shown together with the isotopic composition of natural uranium.
  • Table 2 shows the fission cross section ( ⁇ f ) and capture cross section ( ⁇ a ) of each uranium isotope at thermal neutron energy (0.0253 eV) (nuclear data library JENDL-3.3, Japan Atomic Energy Agency) July 27, 2007). The unit is barn.
  • a fuel assembly for the CANDU furnace shown in FIG. 1 was designed using natural uranium and recovered uranium having the composition shown in Table 1.
  • 11 is a bearing pad
  • 12 is a cladding tube
  • 13 is an end support plate
  • 14 is a fuel pellet loaded in the cladding tube
  • 15 is a spacer
  • 16 Is an end plug.
  • the end support plate 13 is fixed to the end of the fuel assembly 10 by resistance welding, and the end plug 16 is fixed by resistance welding to seal the end of the cladding tube 12.
  • the pad 11 is brazed to the cladding tube 12.
  • the length of the fuel assembly 10 is 49.5 cm, the outer diameter of the cladding tube 12 is 13.61 mm, the wall thickness is 0.419 mm, and the diameter of the fuel pellet 15 is 12.154 mm.
  • the length is 16.40 mm and the density is 10.6 g / cm 3 .
  • the energy distribution (neutron flux spectrum) of neutrons in the fuel due to the fuel difference (natural uranium and recovered uranium) in this fuel assembly was calculated.
  • the result is shown in FIG.
  • the horizontal axis is a logarithmic scale of neutron energy (eV)
  • the vertical axis is the neutron density in the fuel (Source / Letergy / volume).
  • the thick line is for natural uranium
  • the thin line is for recovered uranium.
  • the neutron flux spectrum in PWR is also indicated by a dotted line for reference.
  • the horizontal axis represents the burnup (burnup of the fuel assembly) (MWd / t)
  • the left vertical axis represents the infinite multiplication factor (k-infinity)
  • the right vertical axis represents both fuels.
  • Infinite multiplication factor difference ⁇ k for recovered uranium-for natural uranium (%).
  • the line connecting the black diamonds is natural uranium fuel
  • the line connecting the black squares is the recovered uranium fuel
  • the line connecting the white circles is the difference between the infinite multiplication factors of both fuels.
  • the current CANDU furnace employs 7500 MWd / t as the take-off burnup.
  • the take-off burnup of recovered uranium corresponding to the infinite multiplication factor at this point is about 9500 MWd as shown by the arrow in FIG. / T. This indicates that by using recovered uranium fuel in the CANDU furnace, the fuel can be used over a long period of about 25%, and it becomes possible to reduce the fuel cycle cost and the radioactive waste.
  • the value coefficient of uranium isotope and the value of recovered uranium will be described below.
  • the recovered uranium contains more U235 that easily undergoes fission reaction than natural uranium. As a result, it can be said that the fuel is more valuable than natural uranium.
  • the composition of uranium isotopes in the recovered uranium of the light water reactor does not always follow Table 1, but depends on the enrichment when loaded into the light water reactor and the degree of burnup extracted from the light water reactor.
  • the recovered uranium contains U234 that is only slightly contained in natural uranium and U236 that does not exist, both of which act as poisons that capture neutrons. To do. Therefore, for example, when the proportion of U235 in the recovered uranium is the same as that of natural uranium, depending on the composition of the recovered uranium, the generated U234 and U236 may be less valuable than natural uranium. May be less valuable than uranium.
  • the infinite multiplication factor critical fixed value
  • the infinite multiplication factor is the ratio of the neutron production reaction to the neutron absorption reaction in the reactor, and is 1 in the critical state, and 1 or more is established as a nuclear reactor.
  • the ratio of neutron production and neutron absorption of each uranium isotope in the fuel is calculated using the number of neutrons generated by one fission ( ⁇ ), fission cross section ( ⁇ f ), and capture cross section ( ⁇ a ), ( ⁇ ⁇ ⁇ a / ⁇ f ) (this value is defined as a value coefficient), so that the sum of the product of the value coefficient of each uranium isotope and the composition ratio of the uranium isotope Can be evaluated as value.
  • the value coefficient of the recovered uranium isotope at 0.0253 eV which is the neutron utilization energy is obtained in advance.
  • the recovered uranium is used as a fuel in a heavy water reactor or the like. The value is calculated and compared with the value of separately prepared natural uranium, and it is determined whether or not the recovered uranium is reused as fuel in a heavy water reactor or the like.
  • the value of the recovered uranium is larger than that of natural uranium, it is judged to be valuable as a fuel for heavy water reactors, and if it is smaller, it is not used as a fuel for heavy water reactors. Concentrate or use for MOX fuel production.
  • composition ratio of each uranium isotope in the recovered uranium is obtained from calculation and experience from the composition and combustion history of the spent fuel of the light water reactor that is the raw material of the recovered uranium, and the measurement of radiation from the recovered uranium. The value confirmed by such as is adopted.
  • Table 3 shows the value coefficient ( ⁇ f / ⁇ a ) of each uranium isotope for thermal neutrons (0.0253 eV).
  • ⁇ is the average number of neutrons generated per fission of the uranium isotope.
  • Table 4 compares the values of recovered uranium and natural uranium having the compositions shown in Table 1 as fuel. As shown in Table 4, the value of natural uranium is 1.50 and the value of recovered uranium is 1.73. For this reason, it can be seen that the recovered uranium having the composition shown in Table 1 has a higher value as a fuel than natural uranium, and thus is worth burning in a CANDU furnace. Furthermore, it can be seen that recovered uranium having a value of 1.50 or less has little merit for use as fuel for the CANDU furnace.
  • Example 1 Next, as a specific example, several fuels using recovered uranium under conditions different from those in Table 1 were evaluated using the above-described value coefficient and the like.
  • STEP 1 fuel (concentration: 4.1%) and STEP 2 fuel (concentration: 4.8%) were taken out from PWR at 30 GWd / t, 40 GWd / t, and 50 GWd / t, respectively.
  • a total of 6 types of recovered uranium, and from BWR, 3 types of 8 ⁇ 8 fuel (concentration: 3.6%) were extracted and recovered at an extraction burnup of 30 GWd / t, 40 GWd / t, and 50 GWd / t.
  • Table 5 shows the value of these recovered uranium as fuel.
  • the numerical value enclosed in parentheses (), for example, (35) in the extracted burn-up degree column, is the burn-out burn-out degree of BWR 9 ⁇ 9 fuel.
  • FIG. 4 conceptually shows changes in burnup and infinite multiplication factor when a total of 6 types of recovered uranium recovered from the above 6 types of fuel burned by PWR and natural uranium are burned in a CANDU furnace. (Approximate state).
  • FIG. 5 conceptually shows changes in burnup and infinite multiplication factor when a total of 6 types of uranium recovered from the 6 types of fuel burned in BWR and natural uranium are burned in a CANDU furnace. Show. 4 and 5, the horizontal axis represents the burnup (MWd / t), and the vertical axis represents the infinite multiplication factor (k-infinity).
  • the line indicated by reference numeral 1 in FIG. 4 is for STEP2 fuel when the burnup degree is 30 GWd / t
  • the line indicated by reference numeral 2 is when STEP2 fuel is taken out and the burnup degree is 40 GWd / t
  • STEP1 fuel is taken out with burnup degree Is 30 GWd / t
  • the line indicated by reference numeral 3 is the case where the take-off burnup is 50 GWd / t with STEP2 fuel and the case where the takeoff burnup is 40 GWd / t with STEP1 fuel
  • the line indicated by reference numeral 4 is STEP1
  • the case where the fuel is taken out by fuel and the burnup is 50 GWd / t and the case of natural uranium.
  • the line indicated by reference numeral 1 in FIG. 5 is the case where the take-off burnup is 30 GWd / t with 8 ⁇ 8 fuel and the case where the take-off burnup is 35 GWd / t with 9 ⁇ 9 fuel
  • the line indicated by reference numeral 2 is 8 ⁇ 8 fuel with a take-off burnup of 40 GWd / t, 9 ⁇ 9 fuel with a take-off burnup of 45 GWd / t, and natural uranium.
  • This is a case where the degree is 50 GWd / t and a case where the degree of combustion with 9 ⁇ 9 fuel is 55 GWd / t.
  • the uranium recovered from the spent fuel of PWR has a greater value than the natural uranium.
  • the recovery uranium from the spent fuel of BWR is the ultimate burnup limit of ⁇ 10 GWd / t (40 GWd / t for 8 ⁇ 8 fuel, 45 GWd / t for 9 ⁇ 9 fuel). If it is extracted at a burnup of about t) or less, it can be seen that the value is equivalent to that of natural uranium fuel.
  • Example 2 The present embodiment relates to using spent fuel taken out from a light water reactor for a graphite furnace.
  • natural uranium fuel is used in the same manner as heavy water moderators because graphite, which has a larger reduction ratio than light water, is used as a moderator. Therefore, also in this case, the recovered uranium fuel can be used.
  • the neutron spectrum due to the difference in the structure of the reactor due to the reduction ratio of heavy water and graphite and the difference in physical state, depending on the case, the neutron spectrum, the number of neutrons generated by one fission, from generation to annihilation The fission cross section and capture cross section of neutrons are slightly different. As a result, the value used for evaluating the recovered uranium fuel may be slightly different from the value intended for heavy water reactors.
  • Example 3 The present embodiment relates to burning again in a heavy water reactor or the like without separating plutonium and uranium recovered from fuel used in a light water reactor.
  • the fuel of the present embodiment is more valuable than the fuel using only recovered uranium, and can be used without any problem even in a furnace using micro-enriched uranium.
  • uranium and plutonium differ in nuclear properties such as the number of neutrons generated by fission (plutonium is more than uranium), fission cross section, and neutron capture cross section. However, there are also differences in the neutron spectra in the furnace.
  • Example 2 the basics are the same as in Example 1, and further, the data necessary for the examination when burning the recovered uranium-plutonium mixed fuel in place of the recovered uranium fuel is specifically the nuclear properties of each plutonium isotope nuclide. May be necessary for the development of a fast breeder reactor, and is published in many research institutions, books, etc., for example, in the nuclear data library JENDL-3.3 of the Japan Atomic Energy Agency. Various programs have already been developed, such as calculating the change in infinite multiplication factor of fuel with the progress of combustion of recovered uranium-plutonium fuel for heavy water reactors and graphite reactors.
  • the core is divided into meshes, and nuclear reactions and neutron fluxes per unit time are obtained by calculation within each mesh.
  • the basic calculation process is the same as that for light water reactors, such as repeating the same calculation process using values that have been exchanged and exchanged in the next unit time. It can be done easily.
  • the BWR has a larger core than the PWR and has a high degree of freedom in arranging fuel in the furnace at the start of each cycle operation. For this reason, the fuel taken out from the BWR as used is often in a state close to the maximum value of the ultimate burnup (allowable maximum burnup). As a result, fuel using only uranium recovered from spent BWR fuel is often less valuable than fuel using natural uranium. However, if the fuel uses recovered uranium-plutonium, not only will it be more valuable than natural uranium, but it will use not only CANDU furnaces that burn natural uranium but also micro-enriched uranium with 1-2% U235. It can also be used in a furnace.
  • Example 4 The present embodiment is a modification of the third embodiment, and relates to burning plutonium and uranium extracted from a disassembled nuclear bomb into recovered uranium-plutonium.
  • uranium-plutonium recovered from BWR is generally less valuable than uranium-plutonium recovered from PWR.
  • Pu239 and U235 used for nuclear bombs have a concentration of 90% or more. For this reason, when Pu239 or U235 extracted from a discarded nuclear bomb is slightly mixed with uranium-plutonium recovered from BWR, the value of the fuel increases greatly, and it is optimal for burning in heavy water reactors and graphite reactors. It will be a valuable fuel. And effective use of nuclear bombs and peaceful treatment can be achieved.
  • the present embodiment is also a modification of the third embodiment and relates to burning in a CANDU furnace or the like by focusing on only plutonium in uranium-plutonium recovered from BWR. That is, unlike U235 in the recovered uranium, the concentration of Pu239 in the recovered plutonium, that is, the plutonium isotope that actually undergoes fission is high. For this reason, there is no need to concentrate Pu239 from the recovered plutonium. Thus, for example, when the recovered plutonium is burned in a CANDU furnace, it can be burned by blending with natural uranium or by blending with recovered uranium-plutonium.
  • recovered uranium remains, but the recovered uranium is loaded into a fast breeder reactor and temporarily stored in a storage facility.

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Abstract

La présente invention concerne un combustible de réacteur nucléaire destiné à être utilisé dans un réacteur à eau lourde ou un réacteur à graphite, produit à partir d’une substance combustible récupérée d’un combustible utilisé dans à réacteur à eau légère sans enrichissement en U235 contenu dans la substance combustible. En outre, l’invention concerne un procédé permettant de produire un combustible de réacteur nucléaire destiné à être utilisé dans un réacteur à eau lourde ou un réacteur à graphite, comprenant l’étape consistant à récupérer la substance combustible d’un combustible utilisé dans un réacteur à eau légère et l’étape consistant à produire un combustible à partir de la substance combustible récupérée sans enrichissement en U235 contenu dans la substance combustible.
PCT/JP2008/060569 2008-06-09 2008-06-09 Combustible pour réacteur à eau lourde ou réacteur à graphite et son procédé de production WO2009150710A1 (fr)

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PCT/JP2008/060569 WO2009150710A1 (fr) 2008-06-09 2008-06-09 Combustible pour réacteur à eau lourde ou réacteur à graphite et son procédé de production
CA2724582A CA2724582A1 (fr) 2008-06-09 2008-06-09 Combustible pour reacteur a eau lourde ou reacteur a graphite et son procede de production
KR1020107028123A KR101488555B1 (ko) 2008-06-09 2008-06-09 중수로 또는 흑연로용 연료 및 그 제조방법

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