WO2009150710A1 - 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 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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French (fr)
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/en
Priority to CA2724582A priority patent/CA2724582A1/en
Priority to KR1020107028123A priority patent/KR101488555B1/en
Publication of WO2009150710A1 publication Critical patent/WO2009150710A1/en

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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.

Abstract

A nuclear reactor fuel for use in a heavy-water reactor or graphite reactor, produced from a fuel substance recovered from a fuel spent in a light-water reactor without enrichment of U235 contained in the fuel substance. Further, there is provided a process for producing a nuclear reactor fuel for use in a heavy-water reactor or graphite reactor, comprising the fuel substance recovering step of recovering a fuel substance from a fuel spent in a light-water reactor and the fuel producing step of producing a fuel from the recovered fuel substance without enrichment of U235 contained in the fuel substance.

Description

重水炉または黒鉛炉用燃料及びその製造方法Fuel for heavy water reactor or graphite furnace and method for producing the same
　本発明は重水炉または黒鉛炉用燃料及びその製造方法に関し、特に軽水炉で使用済みとされた燃料から回収された燃料性物質を、前記燃料性物質中に含まれているＵ２３５を濃縮することなく用いて製造される重水炉または黒鉛炉用燃料及びその製造方法に関する。 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.
　我国で広く運転されている軽水を減速材として使用する原子炉（以下、「軽水炉」と記す）では、ＰＷＲ（加圧水型炉）であっても、ＢＷＲ（沸騰水型炉）であっても、燃料としてＵ（ウラン）２３５が３．６～５質量％（以下、単に「％」と記す）程度に濃縮された燃料（以下、「濃縮ウラン」とも記す）が使用されている。具体的には、ＰＷＲではＵ２３５の濃縮度（以下、単に「濃縮度」と記す）が４．１％の燃料（ＳＴＥＰ１燃料）と４．８％の燃料（ＳＴＥＰ２燃料）が用いられており、ＢＷＲでは濃縮度が３．６％程度の燃料（燃料棒を８×８に配列した燃料集合体の場合）と４％程度の燃料（燃料棒を９×９に配列した燃料集合体の場合）が用いられている。 In reactors that use light water as a moderator (hereinafter referred to as “light water reactor”) that is widely operated in Japan, whether it is a PWR (pressurized water reactor) or a BWR (boiling water reactor), A fuel (hereinafter also referred to as “enriched uranium”) 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. Specifically, in PWR, U235 enrichment (hereinafter simply referred to as “enrichment”) is 4.1% fuel (STEP1 fuel) and 4.8% fuel (STEP2 fuel). BWR has a fuel enrichment of about 3.6% (in the case of a fuel assembly with 8 × 8 fuel rods) and 4% fuel (in the case of a fuel assembly with 9 × 9 fuel rods) Is used.
　ＰＷＲ、ＢＷＲとも燃料は所定の燃焼度に到達すれば使用済み燃料として炉内から取り出され、一定期間保存された後再処理に付される。具体的な所定の燃焼度としては、ＰＷＲでは、濃縮度が４．１％の燃料と４．８％の燃料の到達燃焼度の上限（許容最高燃焼度）は各々４８ＧＷｄ／ｔと５５ＧＷｄ／ｔであるが、炉内での燃料の装荷位置により燃焼度が相違し、また運転継続中には炉外へ取出せないため、実際には各々４０ＧＷｄ／ｔと４５ＧＷｄ／ｔ程度の状態で取出されている。また、ＢＷＲでは、濃縮度が３．６％程度の燃料と４．０％程度の燃料の到達燃焼度の上限は各々４８ＧＷｄ／ｔと５５ＧＷｄ／ｔであり、ＢＷＲはＰＷＲに比べて炉心が大きいため炉内で各燃料を装荷する際の配置の自由度が高いため、いずれの燃料もこれらに近い到達燃焼度で取り出されている。 ∙ For both PWR and BWR, 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. As the specific 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. However, 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. In 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.
　なお、参考までに、濃縮度を前記の値としているのは、高すぎると濃縮のコストがかかりすぎること、新燃料を装荷した直後の炉心の反応度が高くなりすぎる（後で説明する無限増倍率が大きくなりすぎる）こと等を考慮したものである。また、到達燃焼度の上限は、燃焼に伴ってＵ２３５が減少し、さらに核***で生じた中性子を吸収する核種が増加してくるため炉心を臨界に保持し難くなること、燃料ペレットや燃料棒の被覆管の損傷を防止する必要があること、等を考慮したものである。 For reference, 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). In addition, 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.
　また、実際には、燃料の炉内位置の検討、炉内への装荷、所定の燃焼度に到達したか否かの判断、炉外への取り出し、炉外での保管等は、多数の燃料棒を所定の構造、形状、寸法に組み立てた燃料集合体を単位として行われる。ただし、これは自明の事項であり、また一々正確に記すと文が煩雑となる。このため、特に区別したり、言及したりする必要があると思われる場合を除いて、単に「燃料」と記す。 Actually, many fuels are used for examining the position of the fuel in the furnace, loading it into the furnace, determining whether the predetermined burnup has been reached, taking it out of the furnace, storing it outside the furnace, etc. The fuel assembly in which the rods are assembled in a predetermined structure, shape, and dimensions is used as a unit. However, this is a self-evident matter, and if written exactly one by one, the sentence becomes complicated. For this reason, it is simply referred to as “fuel” unless there is a particular need for distinction or mention.
　次に、商用炉におけるＵ２３５の核***の連鎖反応の制御について説明する。核***の連鎖反応は、制御の容易さ等を考慮して、核***により発生した直後の高速中性子（２ＭｅＶ程度のエネルギーを持つ）をそのまま用いるのではなく、減速材を用いて一旦熱中性子（０．０２５３ｅＶ程度のエネルギーを持つ）に減速させ、その後Ｕ２３５に吸収させることにより持続される様になされている。また、Ｕ２３５の核***で発生した中性子の中には、減速の途中に炉外に漏れ出したり、減速材や中性子毒やＵ２３８や核***で生じた核種等に吸収されたりするものがあり、これらの核***の連鎖反応に寄与しない中性子の比率は、燃焼の進行に伴い逆に増加していく。このため、実際の原子炉の運転では、燃焼に伴う炉心の反応度の低下を補償するため、種々の対策、調節がなされている。 Next, control of the chain reaction of U235 fission in a commercial reactor will be described. In the fission chain reaction, considering the ease of control, fast neutrons (having energy of about 2 MeV) immediately after the fission is not used as they are, but thermal neutrons (0. It has an energy of about 0253 eV) and then is absorbed by U235 so as to be sustained. Also, some neutrons generated by U235 fission may leak out of the reactor during deceleration, or may be absorbed by moderators, neutron poisons, U238 or nuclides produced by fission, etc. The ratio of neutrons that do not contribute to the fission chain reaction increases conversely with the progress of combustion. For this reason, in actual operation of the nuclear reactor, various countermeasures and adjustments are made to compensate for a decrease in the reactivity of the core due to combustion.
　例えば、軽水炉においては、ＰＷＲとＢＷＲでは多少の相違があるが、燃焼に伴って冷却材中のボロン等の中性子毒の濃度を下げたり、制御棒の挿入を少なくしたりする、１年程度継続する各運転サイクルの終了後に次の運転サイクルでも燃やす各燃料（集合体）を炉内に装荷する位置を、その燃焼度を考慮して決める、燃料中に可燃性毒物を配合しておく等である。また、天然ウランや微濃縮ウランを用いる原子炉では、一般的に軽水炉に比較して炉心が大きく、さらに炉心は鋼鉄製の圧力容器に格納されていないため、運転継続中に所定の燃焼度に達した燃料を炉外へ取り出し、それに換えて新燃料を装荷することがなされている。 For example, in light water reactors, there is a slight difference between PWR and BWR, but the concentration of neutron poisons such as boron in the coolant decreases with combustion, and the insertion of control rods continues for about one year. After the end of each operation cycle, determine the position where each fuel (aggregate) to be burned in the next operation cycle is loaded in the furnace in consideration of its burnup, such as by adding a flammable poison in the fuel is there. In addition, reactors using natural uranium or microenriched uranium generally have a larger core than light water reactors, and the core is not stored in a steel pressure vessel. The reached fuel is taken out of the furnace and replaced with new fuel.
　我国では商用炉は軽水炉しかないが、世界的には、カナダを中心として、Ｕ２３５が０．７２％しか含まれていない天然ウランや１％～２％に濃縮した微濃縮ウランを燃料として用いるタイプの原子炉、具体的には重水を減速材として用いる炉（例えば、ＣＡＮＤＵ炉）や、その他黒鉛を減速材として用いるタイプの炉が商用炉として稼動している。この様な原子炉が用いられる理由は、重水の減速比（１回の衝突当りのエネルギー対数の平均の減少率×散乱断面積／吸収断面積）は軽水の８０倍、黒鉛は同じく２．４倍もあり、このため重水を使用するＣＡＮＤＵ炉等では天然ウランでも炉心を臨界に保持することが可能となり、高価な濃縮ウランを燃料に使用する必要がないことによる。 In Japan, 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. In particular, a reactor using heavy water as a moderator (for example, a CANDU furnace), and 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.
　ただし、天然ウランや微濃縮ウランを用いると、核***をおこすＵ２３５の含有量が本来的に少ないため、炉内における燃焼に伴って、炉心の反応度余裕が急速に少なくなり、長期間にわたって炉心を臨界に維持することは困難である。このため、燃料の取り出し燃焼度は、軽水炉に比較してはるかに小さくなる。例えば、カナダ、その他韓国等世界７カ国で合計３２基運転されているＣＡＮＤＵ炉においては、７．５ＧＷｄ／ｔ程度である。 However, if natural uranium or slightly enriched uranium is used, the content of U235 that causes fission is inherently low, so the reactivity margin of the core decreases rapidly with the combustion in the reactor, and the core remains for a long period of time. It is difficult to maintain criticality. For this reason, 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.
　なお、天然ウランや微濃縮ウランを用いる原子炉、例えば重水を用いるＣＡＮＤＵ炉においては、軽水炉で使用される様な濃縮ウランを装荷して取り出し燃焼度を増大させることは実用化されていない。その理由は、濃縮ウランは高価であり、また濃縮ウランそのままでは原子炉の反応度、特に装荷直後の反応度が高くなりすぎるため安全運転が困難となり、その対策として濃縮ウランに中性子毒物を多量に配合して反応度を低下させれば中性子経済が悪化することによる。 Incidentally, in a nuclear reactor using natural uranium and fine enriched uranium, for example, a CANDU reactor using heavy water, it is not put into practical use to load and take out enriched uranium as used in a light water reactor to increase the burnup. The reason for this is that enriched uranium is expensive, and if the enriched uranium is used as it is, the reactivity of the reactor, especially the reactivity immediately after loading becomes too high, making safe operation difficult. This is because the neutron economy deteriorates if it is mixed to lower the reactivity.
　次に、軽水炉であれ、天然ウランや微濃縮ウランを用いる原子炉であれ、所定の燃焼度に達したため使用済み燃料として炉外に取り出された燃料は、Ｕ２３５の核***で生じた放射性元素（娘核種）から強烈な放射が放出され、また発熱もあるため、所定の設備を有する貯蔵用プール内で一定期間保管され、その後燃え残ったウランを回収したり、ウランの燃焼で生じたプルトニウムを回収したりする等の再処理がなされることとなる（非特許文献１）。
清瀬量平著、「使用済み燃料とプルトニウムの化学工業」、日刊工業新聞社刊、１９８４年 Next, whether it is a light water reactor or a nuclear reactor that uses natural uranium or micro-enriched uranium, 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
　しかしながら、近年の原油価格の高騰の下、天然ウランや微濃縮ウランを用いる原子炉及び軽水炉双方において、燃料サイクルコストの低下、資源利用率の向上（即ち、資源の有効利用）、放射性廃棄物の減少に対する要請は益々厳しくなってきている。 However, under the recent rise in crude oil prices, in both nuclear reactors and light water reactors using natural uranium and microenriched uranium, fuel cycle costs are reduced, resource utilization is improved (ie, effective use of resources), and radioactive waste The demand for decline has become increasingly severe.
　このため、天然ウランや微濃縮ウランを用いる原子炉のみならず軽水炉においても、燃料サイクルコストが低下し、資源利用率が向上し、放射性廃棄物が減少する技術の開発が望まれていた。 For this reason, it has been desired to develop a technology for reducing the fuel cycle cost, improving the resource utilization rate, and reducing the radioactive waste not only in reactors using natural uranium and fine enriched uranium but also in light water reactors.
　本発明は、以上の課題を解決することを目的としてなされたものであり、軽水炉で使用済みとされた燃料を天然ウランや微濃縮ウランを用いる原子炉で燃やす様にしたものである。以下、各請求項の発明を説明する。 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.
　請求項１に記載の発明は、
　重水炉または黒鉛炉に用いられる原子炉用燃料であって、
　軽水炉で使用済みとされた燃料から回収された燃料性物質を、前記燃料性物質中に含まれているＵ２３５を濃縮することなく用いて製造されていることを特徴とする原子炉用燃料である。 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. .
　本請求項の発明においては、軽水炉で使用済みとされた燃料から回収された燃料性物質を、その燃料性物質中に含まれているＵ２３５を濃縮することなく、本来は天然ウランや微濃縮ウランを燃料として用いる重水炉または黒鉛炉で再度燃やすため、重水炉や黒鉛炉においては、燃料サイクルコストの低下、資源利用率の向上、放射性廃棄物の減少を図ることができる。一方、軽水炉においても、濃縮費の削減や濃縮プロセスの不要による燃料サイクルコストの低下、資源の有効利用、放射性廃棄物の減少を図ることができる。 In the invention of the present claim, 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. On the other hand, 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.
　ここに、「使用済みとされた燃料」とは、到達燃焼度の上限やそれに近い燃焼度に到達した燃料のみならず、もう一度次のサイクル運転で燃やそうとしても、当該次のサイクル運転の途中で許容最高燃焼度に到達するため、結果的に次のサイクル運転で燃やすことができない燃料をも含む。 Here, 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.
　また、「燃料」とは、現在の軽水炉は主に天然ウランを濃縮したウラン燃料が用いられているが、将来的にはプルトニウムを配合したＭＯＸ燃料を用いることが計画されている。また、将来的にはトリウム燃料も開発中である。このため、ウラン燃料に限定されるものではない。 As the “fuel”, 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.
　また、「燃料性物質」とは、核***性物質及びその同位体を指し、現在の我国の軽水炉では、Ｕ２３５とＰｕ２３９及びそれらの同位体であるが、将来的に可能性があるトリウムを排除するものではなく、さらに不可避的に含有され得るその他の不純物元素を含んでいてもよい。 “Fuel 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.
　また、「前記燃料性物質中に含まれているＵ２３５を濃縮することなく用いて製造されている」とは、使用済みとされた燃料から回収されたウラン中のＵ２３５を濃縮しない限り、使用済み燃料からウランとプルトニウムを分離せずにＰｕ２３９等をも核***性物質として用いて製造される場合も含む。この場合、さらに核爆弾から回収した核***性物質または天然ウランや天然ウランを微濃縮したウランを配合して製造されている場合も含む。
　また、回収された燃料性物質からウランを除き、プルトニウムのみを用いて製造される場合等も含む。 In addition, “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. This includes cases where Pu239 or the like is also used as a fissile material without separating uranium and plutonium from the fuel. In this case, it also includes the case where it is manufactured by blending fissile material recovered from nuclear bombs or natural uranium or uranium slightly enriched with natural uranium.
In addition, it includes cases where uranium is removed from the recovered fuel substance and the product is produced using only plutonium.
　また、「Ｕ２３５を濃縮することなく用いて製造」とは、重水炉または黒鉛炉用燃料を製造する際に、回収ウラン中のＵ２３５の濃縮を行わないことを指す（天然ウランと異なり、回収ウランには強度の放射能源となるＵ２３２が含まれるため、濃縮工程が複雑となり、コストもかかる）。 In addition, “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).
　また、原子炉用燃料集合体を構成する燃料棒中の燃料ペレットも本発明の「原子炉用燃料」に該当し、たとえ燃料ペレットが１個の場合でも本発明の「原子炉用燃料」に該当する。 Further, 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.
　なお、「重水炉」、「黒鉛炉」とは、重水または黒鉛を減速材として用いる限り、ガスや軽水を冷却材として使用する原子炉も含む。 Note that 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.
　請求項２に記載の発明は、
　前記軽水炉で使用済みとされた燃料から回収された燃料性物質が、ウランであることを特徴とする請求項１に記載の原子炉用燃料である。 The invention described in claim 2
The fuel for a nuclear reactor according to claim 1, wherein the fuel material recovered from the fuel used in the light water reactor is uranium.
　本請求項の発明においては、重水炉または黒鉛炉用の原子炉用燃料が、軽水炉で使用済みとされた燃料から回収されたウランを用いて製造されているため、軽水炉で燃え残ったＵ２３５を主に、その他微量のウラン同位体が重水炉または黒鉛炉で燃やされることとなり、重水炉または黒鉛炉のみならず軽水炉にとっても燃料コストの低下になる。 In the invention of this claim, since 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.
　また、軽水炉で使用済みとされた燃料には、核***性物質であるＵ２３５が天然ウランに比べ多く含まれているため、請求項１の発明の効果を特に良好に発揮できることとなる。 Further, since 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.
　請求項３に記載の発明は、
　前記軽水炉が、ＰＷＲであることを特徴とする請求項１または請求項２に記載の原子炉用燃料である。 The invention described in claim 3
3. The nuclear fuel according to claim 1, wherein the light water reactor is a PWR.
　本請求項の発明においては、請求項１または請求項２の発明の効果を、一層良好に発揮できる。即ち、ＰＷＲはＢＷＲに比べて炉心が小さく、さらに炉内における燃料の配置の自由度が少ないため、前記した様に、許容最高燃焼度にはある程度の余裕をもった状態で使用済みとせざるを得ない燃料が多く発生する。その結果、使用済みとされた燃料中のＵ２３５の組成比は天然ウランにおける組成比よりも高く、その分重水炉や黒鉛炉に用いた場合、燃焼度を高くでき、燃料交換のインターバルを長くすることができるからである。 In the invention of this claim, the effect of the invention of claim 1 or claim 2 can be exhibited more satisfactorily. In other words, 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.
　請求項４に記載の発明は、
　前記原子炉用燃料が、ＣＡＮＤＵ炉用の燃料であることを特徴とする請求項１または請求項２に記載の原子炉用燃料である。 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.
　また、請求項５に記載の発明は、
　前記原子炉用燃料が、ＣＡＮＤＵ炉用の燃料であることを特徴とする請求項３に記載の原子炉用燃料である。 The invention according to claim 5
The reactor fuel according to claim 3, wherein the reactor fuel is a fuel for a CANDU reactor.
　これらの請求項の発明においては、請求項１から請求項３のいずれか１項の発明の効果が最も良く発揮できる。即ち、ＣＡＮＤＵ炉は減速材として減速比が大きい重水を用いるため、燃料中の核***性物質の組成比が相当小さくなっても炉心を臨界に保つことが可能である。このため、軽水炉で最高燃焼度に到達した燃料であっても、それから回収した燃料性物質を用いて製造したＣＡＮＤＵ炉用の燃料は、天然ウランを用いた燃料に比較して充分大きな燃焼度を得られることが多いからである。 In the inventions of these claims, the effect of the invention of any one of claims 1 to 3 can best be exhibited. That is, since the CANDU furnace uses heavy water having a large reduction ratio as a moderator, the core can be kept critical even if the composition ratio of the fissile material in the fuel is considerably reduced. For this reason, even if the fuel reaches the maximum burnup in the light water reactor, the fuel for the CANDU furnace manufactured using the fuel material recovered from it has a sufficiently high burnup compared to the fuel using natural uranium. This is because it is often obtained.
　請求項６に記載の発明は、
　重水炉または黒鉛炉に用いられる原子炉用燃料の製造方法であって、
　軽水炉で使用済みとされた燃料から燃料性物質を回収する燃料性物質回収ステップと、
　前記回収された燃料性物質を、前記燃料性物質に含まれているＵ２３５を濃縮することなく用いて燃料を製造する燃料製造ステップと
を有していることを特徴とする原子炉用燃料の製造方法である。 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.
　本請求項の発明は、請求項１の発明を方法の発明として捉えたものである。 The invention of this claim captures the invention of claim 1 as a method invention.
　請求項７に記載の発明は、
　前記燃料性物質回収ステップの後に、
　前記回収された燃料性物質中の各組成物質の組成比を求める組成比算出ステップと、
　前記各組成物質毎に、核***断面積と核***で生じる中性子の個数との積を捕獲断面積で割り、その商を以て各組成物質の価値係数とする価値係数算出ステップと、
　前記各組成物質毎に、前記組成比と前記価値係数の積を求め、その総和を以て前記回収された燃料性物質の価値とする価値算出ステップと、
　前記燃料性物質の価値と所定の基準値とを比較し、その大小に応じて、前記燃料製造ステップへの移行を判断する再利用判断ステップと
を有していることを特徴とする請求項６に記載の原子炉用燃料の製造方法である。 The invention described in claim 7
After the fuel substance recovery step,
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. A method for producing a fuel for a reactor as described in 1).
　本請求項の発明においては、回収された燃料性物質の価値と所定の基準値とを比較して、当該燃料性物質を重水炉や軽水炉用燃料として用いるか否かを判断しているため、判断の結果、軽水炉で使用済みとされた燃料を用いて製造された燃料を、重水炉または黒鉛炉に用いた場合、到達燃焼度が大きくなり、燃料交換のインターバルを長くすることができる。 In the invention of the present claim, since 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. As a result of the determination, when a fuel manufactured using a fuel that has been used in a light water reactor is used in a heavy water reactor or a graphite reactor, the burn-up degree is increased, and the fuel exchange interval can be extended.
　ここに、厳密には、核***断面積（σｆ）、核***で生じる中性子の個数（ν）、捕獲断面積（σａ）は中性子エネルギーに依存するが、本発明においては、対象がＣＡＮＤＵ炉といった熱中性子炉であり、炉内で核反応に寄与する中性子のほとんどが熱中性子のため、熱中性子エネルギー（０．０２５３ｅＶ）における各値を用いて、下記式により価値係数を算出する。
　　　　価値係数＝ν×σａ／σｆ Strictly speaking, the fission cross section (σ f ), the number of neutrons generated by fission (ν), and the capture cross section (σ a ) depend on the neutron energy. In the present invention, the object is a CANDU reactor. Since it is a thermal neutron reactor and most of the neutrons contributing to the nuclear reaction in the reactor are thermal neutrons, the value coefficient is calculated by the following formula using each value in thermal neutron energy (0.0253 eV).
Value coefficient = ν × σ a / σ f
　また、「回収された燃料性物質中の各組成物質の組成比」とは、ウランとプルトニウムの混合物中におけるウラン同位体とプルトニウム同位体の組成比であっても良いし、ウラン毎、プルトニウム毎の組成比であっても良い。さらに、不可避的に発生したり混入したりした他の原子や元素の組成比が含まれていても良い。組成比の算出については、計算、実測、経験等その手段を問わない。 The “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. A composition ratio of Furthermore, the composition ratio of other atoms and elements that are inevitably generated or mixed may be included. Regarding the calculation of the composition ratio, any means such as calculation, actual measurement, and experience may be used.
　また、「所定の基準値」とは、当該燃料性物質が燃やされる原子炉で本来燃やされる筈の燃料性物質を、例えばＣＡＮＤＵ炉であれば天然ウランを、対象にして各組成物質の組成比、核***断面積、核***で生じる中性子の個数、捕獲断面積から本請求項の発明における燃料性物質の価値算出ステップと同じ手順で求めた値であっても良いし、その値に対して原子炉の型ひいては中性子スペクトル等の相違、組成比の測定等の誤差、例えばＣＡＮＤＵ炉用の燃料とするためのコスト等を反映した修正値であっても良い。 In addition, 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. As a result, 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.
　さらに、「回収された燃料性物質」に核爆弾から取出したプルトニウムやＵ２３５を配合する様な場合には、前記所定の基準値には、その分を見込んで作成されているのは当然である。なおこの場合には、「回収された燃料性物質」と「核爆弾から取出して配合したプルトニウムやＵ２３５」からなる燃料性物質の価値を算出して、所定の基準値と比較する様にしても良い（本発明に均等な発明であり、かかる場合も本請求項の発明に含まれる）。また、別途微濃縮されたＵ２３５を配合する様な場合も同様である。 Furthermore, in the case where plutonium extracted from a nuclear bomb or U235 is blended into the “recovered fuel substance”, it is natural that the predetermined reference value is prepared by taking that amount into account. . In this case, 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). The same applies to the case of blending U235 that is separately finely concentrated.
　さらに、重水炉と黒鉛炉用等の複数の基準値を作成し、その値に応じて再使用する原子炉を変えても良い。 Furthermore, 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.
　請求項８に記載の発明は、
　重水炉または黒鉛炉に用いられる原子炉用燃料の製造方法であって、
　軽水炉で使用済みとされた燃料からウランを回収するウラン回収ステップと、
　前記回収されたウランを、その組成物として含まれているＵ２３５を濃縮することなく用いて燃料を製造する燃料製造ステップと
を有していることを特徴とする原子炉用燃料の製造方法である。 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. .
　本請求項の発明は、請求項２の発明を方法の発明として捉えたものである。 The invention of this claim captures the invention of claim 2 as a method invention.
　請求項９に記載の発明は、
　前記ウラン回収ステップの後に、
　前記回収されたウラン中の各ウラン同位体の組成比を求める組成比表算出ステップと、
　前記各ウラン同位体毎に、核***断面積と核***で生じる中性子の個数の積を捕獲断面積で割り、その商を以て各ウラン同位体の価値係数とする価値係数算出ステップと、
　前記各ウラン同位体毎に、前記組成比と前記価値係数の積を求め、その総和を以て前記回収されたウランの価値とする価値算出ステップと、
　前記回収ウランの価値と所定の基準値とを比較し、その大小に応じて、前記燃料製造ステップへの移行を判断する再利用判断ステップと
を有していることを特徴とする請求項８に記載の原子炉用燃料の製造方法である。 The invention described in claim 9
After the uranium recovery step,
A composition ratio table calculating step for obtaining a composition ratio of each uranium isotope in the recovered uranium;
For each of the uranium isotopes, the product of the fission cross section and the number of neutrons generated by fission is divided by the capture cross section, and the value coefficient calculating step for setting the quotient as the value coefficient of each uranium isotope,
For each of the uranium isotopes, obtain a product of the composition ratio and the value coefficient, and calculate the value of the value of the recovered uranium with the sum of the products,
9. 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.
　本請求項の発明においては、回収されたウランの価値と所定の基準値とを比較して、当該燃料性物質を重水炉や軽水炉用燃料として用いるか否かを判断しているため、判断の結果、軽水炉で使用済みとされた燃料から回収されたウランを用いて製造された燃料を、重水炉または黒鉛炉に用いた場合、到達燃焼度が大きくなり、燃料交換のインターバルを長くすることができる。 In the claimed invention, 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. As a result, when fuel produced using uranium recovered from fuel that has been used in light water reactors is used in heavy water reactors or graphite reactors, the ultimate burnup will increase and the fuel change interval may be lengthened. it can.
　なお、「ウラン」とは、不可避的に他の元素が含まれている場合を含む。またこのため、回収ウランの価値を求める際に、かかる不純物元素の及ぼす影響を経験や実測等に基づいて修正しても良い。 In addition, “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.
　請求項１０に記載の発明は、
　前記軽水炉はＰＷＲであることを特徴とする請求項６ないし請求項９のいずれか１項に記載の原子炉用燃料の製造方法である。 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.
　本請求項の発明においては、請求項６ないし請求項９のいずれか１項の発明の効果を、一層良好に発揮できる。 In the invention of this claim, the effect of the invention of any one of claims 6 to 9 can be exhibited more satisfactorily.
　請求項１１に記載の発明は、
　前記重水炉または黒鉛炉とはＣＡＮＤＵ炉であることを特徴とする請求項６ないし請求項９のいずれか１項に記載の原子炉用燃料の製造方法である。 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.
　また、請求項１２に記載の発明は、
　前記重水炉または黒鉛炉とはＣＡＮＤＵ炉であることを特徴とする請求項１０に記載の原子炉用燃料の製造方法である。 Further, 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.
　これらの請求項の発明においては、請求項６ないし請求項１０のいずれか１項の発明の効果が最も良く発揮できる。 In the inventions of these claims, the effect of the invention of any one of claims 6 to 10 can be exhibited best.
　本発明においては、軽水炉で使用済みとされた燃料を重水炉または黒鉛炉用燃料として燃やすため、天然ウランや微濃縮ウランを用いる原子炉のみならず軽水炉において、燃料サイクルコストの低下、資源利用率の向上、放射性廃棄物の減少を図ることができる。 In the present invention, 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.
ＣＡＮＤＵ炉用の燃料集合体の外観を、概念的に示す図である。It is a figure which shows notionally the external appearance of the fuel assembly for CANDU furnaces. ＣＡＮＤＵ炉とＰＷＲにおける中性子束スペクトル（中性子エネルギー分布）を示す図である。It is a figure which shows the neutron flux spectrum (neutron energy distribution) in a CANDU furnace and PWR. ＣＡＮＤＵ炉において、天然ウランと回収ウランの燃焼に伴う無限増倍率の変化を示す図である。It is a figure which shows the change of an infinite multiplication factor accompanying combustion of natural uranium and recovery uranium in a CANDU furnace. ＰＷＲで燃やした燃料から回収した６種の回収ウランと天然ウランを、ＣＡＮＤＵ炉で燃やした場合の燃焼度と無限増倍率の変化を概念的に示す図である。It is a figure which shows notionally the change of the burnup degree and infinite multiplication factor at the time of burning 6 types of collection | recovery uranium and natural uranium which were collect | recovered from the fuel burned by PWR with a CANDU furnace. ＢＷＲで燃やした燃料から回収した６種の回収ウランと天然ウランを、ＣＡＮＤＵ炉で燃やした場合の燃焼度と無限増倍率の変化を概念的に示す図である。It is a figure which shows notionally the change of a burnup and an infinite multiplication factor at the time of burning 6 types of collection | recovery uranium and natural uranium recovered from the fuel burned by BWR with a CANDU furnace.
符号の説明Explanation of symbols
１１　　ベアリング・パッド
１２　　被覆管
１３　　エンドサポートプレート
１４　　燃料ペレット
１５　　スペーサ
１６　　端栓 11 Bearing pad 12 Clad tube 13 End support plate 14 Fuel pellet 15 Spacer 16 End plug
　以下、本発明の実施の形態につき図を用いて説明する。なお、本発明は以下の実施の形態に限定されるものではない。本発明とは同一及び均等の範囲内において、以下の実施の形態に対して種々の変更を加えることが可能である。 Hereinafter, embodiments of the present invention will be described with reference to the drawings. Note that the present invention is not limited to the following embodiments. Various modifications can be made to the following embodiments within the same and equivalent scope as the present invention.
（ＣＡＮＤＵ炉における回収ウランの利用について）
　本実施例１は、ＰＷＲで使用済みとされた燃料から回収されたウランを、ＣＡＮＤＵ炉で天然ウランに換えて燃やすことに関する。以下、図表を参照しつつ説明する。 (Use of recovered uranium in the CANDU furnace)
The first embodiment relates to burning uranium recovered from fuel used in PWR in place of natural uranium in a CANDU furnace. Hereinafter, description will be made with reference to the drawings.
　表１に、電気出力が１０００ＭＷｅのＰＷＲにおいて所定の燃焼度に到達したため、使用済み燃料として炉外に取り出され、さらに１５０日経過した燃料（ウラン燃料）から回収されたウラン、即ち回収ウランの同位体の組成（％）を、天然ウランにおける同位体の組成と併せて示す。 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.
Figure JPOXMLDOC01-appb-T000001
Figure JPOXMLDOC01-appb-T000001
　表２に、熱中性子エネルギー（０．０２５３ｅＶ）における各ウラン同位体の核***断面積（σｆ）と捕獲断面積（σａ）を示す（核データライブラリＪＥＮＤＬ－３．３、日本原子力研究開発機構、２００７年７月２７日）。なお、単位はバーン（ｂａｒｎ）である。 Table 2 shows the fission cross section (σ f ) and capture cross section (σ 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.
Figure JPOXMLDOC01-appb-T000002
Figure JPOXMLDOC01-appb-T000002
　表１に示す組成の天然ウラン及び回収ウランを用いて、図１に示すＣＡＮＤＵ炉用の燃料集合体を設計した。図１において、１１はベアリング・パッドであり、１２は被覆管であり、１３はエンドサポートプレートであり、１４は被覆管１２内に装荷されている燃料ペレットであり、１５はスペーサであり、１６は端栓である。また、エンドサポートプレート１３は燃料集合体１０の端部に抵抗溶接で固定されており、端栓１６は被覆管１２の端部を密封するために抵抗溶接で固定されており、スペーサ１５とベアリング・パッド１１は被覆管１２にロウ付けされている。 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. In FIG. 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 12, 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.
　また、燃料集合体１０の長さは４９．５ｃｍであり、被覆管１２の外径は１３．０６１ｍｍであり、肉厚は０．４１９ｍｍであり、燃料ペレット１５の直径は１２．１５４ｍｍであり、長さは１６．４０ｍｍであり、密度は１０．６ｇ／ｃｍ３である。 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 .
　次に、この燃料集合体の燃料の相違（天然ウラン及び回収ウラン）による燃料内の中性子のエネルギー分布（中性子束スペクトル）を計算した。その結果を図２に示す。図２において、横軸は中性子のエネルギー（ｅＶ）の対数目盛りであり、縦軸は燃料内の中性子密度（Ｓｏｕｒｃｅ／Ｌｅｔｈａｒｇｙ／体積）である。また、太線は天然ウランの場合であり、細線は回収ウランの場合である。なお、図２には、参考としてＰＷＲにおける中性子束スペクトルをも点線で示してある。 Next, 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. In FIG. 2, the horizontal axis is a logarithmic scale of neutron energy (eV), and the vertical axis is the neutron density in the fuel (Source / Letergy / volume). The thick line is for natural uranium, and the thin line is for recovered uranium. In FIG. 2, the neutron flux spectrum in PWR is also indicated by a dotted line for reference.
　図２において、おおよそ０．３×１Ｅ－２ｅＶから０．３×１Ｅ－１ｅＶの領域で僅かに太線と細線を区別することが可能であるが、他の領域では太線と細線が完全に一致している。即ち、天然ウランと回収ウランとでは、ＣＡＮＤＵ炉内における中性子スペクトルは全中性子エネルギー領域を通じてほとんど相違せず、この結果より、回収ウランを、そのままＣＡＮＤＵ炉で燃やすことが可能であることが判る。なお、エネルギーが０．３×１Ｅ－２ｅＶから０．３×１Ｅ－１ｅＶの領域でＣＡＮＤＵ炉がＰＷＲに比べて中性子束が大であるのは、ＣＡＮＤＵ炉は炉心が大きいため中性子の漏れが少なく、減速材の重水の減速比が大きいことによると推測される。 In FIG. 2, it is possible to distinguish the thick and thin lines slightly in the region of approximately 0.3 × 1E-2eV to 0.3 × 1E-1eV, but in the other regions, the thick line and the thin line completely match. ing. That is, natural uranium and recovered uranium have almost no different neutron spectra in the CANDU furnace throughout the entire neutron energy region, and it can be seen from this result that recovered uranium can be burned in the CANDU furnace as it is. Note that the CANDU reactor has a larger neutron flux than the PWR in the energy range of 0.3 × 1E-2eV to 0.3 × 1E-1eV. It is presumed that the reduction ratio of heavy water in the moderator is large.
　次いで、各燃料による燃焼計算を実施し、燃焼に伴う無限増倍率の変化を比較した。結果を図３に示す。図３において、横軸は燃焼度（燃料集合体の燃焼度）（ＭＷｄ／ｔ）であり、左の縦軸は無限増倍率（ｋ－ｉｎｆｉｎｉｔｙ）であり、右の縦軸は両方の燃料の無限増倍率の差△ｋ（回収ウランの場合－天然ウランの場合）（％）である。また、図中、黒い菱形を結ぶ線は天然ウラン燃料であり、黒い四角を結ぶ線は回収ウラン燃料であり、白丸を結ぶ線は両方の燃料の無限増倍率の差である。 Next, the combustion calculation by each fuel was performed, and the change of the infinite multiplication factor accompanying the combustion was compared. The results are shown in FIG. In FIG. 3, the horizontal axis represents the burnup (burnup of the fuel assembly) (MWd / t), the left vertical axis represents the infinite multiplication factor (k-infinity), and the right vertical axis represents both fuels. Infinite multiplication factor difference Δk (for recovered uranium-for natural uranium) (%). In the figure, the line connecting the black diamonds is natural uranium fuel, the line connecting the black squares is the recovered uranium fuel, and the line connecting the white circles is the difference between the infinite multiplication factors of both fuels.
　現行のＣＡＮＤＵ炉では、取出燃焼度として７５００ＭＷｄ／ｔを採用しているが、この時点での無限増倍率に相当する回収ウランの取出燃焼度は、図３中の矢印で示す様に、約９５００ＭＷｄ／ｔである。これは、ＣＡＮＤＵ炉において、回収ウラン燃料を用いることにより、約２５％程度長期にわたり燃料を使用できることを示しており、燃料サイクルコストの低下と放射性廃棄物の減少を図ることが可能となる。 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.
（回収ウランの価値について）
　次いで、ウラン同位体の価値係数と回収ウランの価値について、以下に説明する。
　表１においては、回収ウランには核***反応をしやすいＵ２３５が天然ウランよりも多く含まれており、結果として天然ウランよりも価値のある燃料ということができる。しかし、軽水炉の回収ウランにおけるウラン同位体の組成は、常に表１に従うわけではなく、軽水炉に装荷された際の濃縮度及び軽水炉から取り出される取出燃焼度に依存している。 (About the value of recovered uranium)
Next, the value coefficient of uranium isotope and the value of recovered uranium will be described below.
In Table 1, 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. However, 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.
　そして、表１に示す様に、回収ウランには、天然ウランでは僅かにしか含まれていないＵ２３４や存在しないＵ２３６が生成されて含まれており、これらはいずれも中性子を捕獲する毒物の働きをする。そのため、例えば、回収ウラン中のＵ２３５の割合が天然ウランと同じ場合には、生成したＵ２３４及びＵ２３６により、天然ウランよりも価値が小さいものとなってしまう様に、回収ウランの組成によっては、天然ウランよりも価値が小さくなる可能性がある。 And, as shown in Table 1, 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.
　基本的には、重水炉や黒鉛炉の様に天然ウラン装荷型原子炉において、無限増倍率（臨界固定値）が天然ウランよりも寿命を通じて大きければ、回収ウランに価値があるということができる。無限増倍率は、炉内における中性子の生成反応と中性子の吸収反応の比であり、臨界状態では１になり、１以上で原子炉として成立する。燃料中の各ウラン同位体の中性子生成と中性子吸収の比は、一回の核***により発生する中性子の個数（ν）、核***断面積（σｆ）、捕獲断面積（σａ）を用いて、（ν×σａ／σｆ）により求めることができる（この値を価値係数と定義する）ため、各ウラン同位体の価値係数と当該ウラン同位体の組成比との積の合計を以て、燃料の価値として評価することができる。 Basically, in a natural uranium loaded nuclear reactor, such as a heavy water reactor or a graphite reactor, if the infinite multiplication factor (critical fixed value) is larger throughout the life than natural uranium, it can be said that recovered uranium is valuable. 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 (ν), fission cross section (σ f ), and capture cross section (σ a ), (Ν × σ a / σ 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.
　具体的には、回収ウランに含まれる各ウラン同位体に対して、例えば、ＣＡＮＤＵ炉であれば、中性子利用エネルギーである０．０２５３ｅＶにおける当該回収ウラン同位体の価値係数を予め求めておく。そして、回収ウランについて、分析により得られたこの回収ウランの各ウラン同位体の組成比と当該ウラン同位体の価値係数との積を基に、当該回収ウランについて重水炉等で燃料として用いる際の価値を算出し、別途作成した天然ウランの価値と比較し、当該回収ウランを重水炉等に燃料として再使用するか否かを判断する。即ち、当該回収ウランの価値が天然ウランより大きければ、重水炉等の燃料として価値があると判断して再使用し、小さければ重水炉等の燃料としては使用せず、他の用途、例えば再度濃縮したり、ＭＯＸ燃料の製造に使用したりする。 Specifically, for each uranium isotope contained in the recovered uranium, for example, in the case of a CANDU furnace, the value coefficient of the recovered uranium isotope at 0.0253 eV which is the neutron utilization energy is obtained in advance. Based on the product of the uranium isotope composition ratio of the recovered uranium obtained by analysis and the value coefficient of the uranium isotope, 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. That is, if 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.
　なお、回収ウラン中の各ウラン同位体の組成比は、当該回収ウランの原料となった軽水炉の使用済み燃料の組成や燃焼履歴等から計算や経験で求め、さらに当該回収ウランからの放射線の測定等で確認を行った値を採用する。 The 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.
　表３に、各ウラン同位体の熱中性子（０．０２５３ｅＶ）に対する価値係数（νσｆ／σａ）を示す。なお、表３におけるνは、当該ウラン同位体の核***一回当たりで発生する中性子の平均個数である。 Table 3 shows the value coefficient (νσ f / σ a ) of each uranium isotope for thermal neutrons (0.0253 eV). In Table 3, ν is the average number of neutrons generated per fission of the uranium isotope.
Figure JPOXMLDOC01-appb-T000003
Figure JPOXMLDOC01-appb-T000003
　表４に、表１に示した組成の回収ウランと天然ウランの燃料としての価値を比較して示す。表４に示す様に、天然ウランの価値は１．５０であり、回収ウランの価値は１．７３である。このため、表１に示した組成の回収ウランは、燃料として天然ウランよりも価値が高いことが、ひいてはＣＡＮＤＵ炉で燃やす価値があることが判る。さらに、価値が１．５０以下の回収ウランは、ＣＡＮＤＵ炉の燃料として用いるメリットが少ないことも判る。 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.
Figure JPOXMLDOC01-appb-T000004
Figure JPOXMLDOC01-appb-T000004
（実施例１）
　次に、具体的な実施例として、前記の価値係数等を用いて、表１と異なる条件の回収ウランを用いた燃料を幾つか評価した。評価は、ＰＷＲからは、ＳＴＥＰ１燃料（濃縮度４．１％）及びＳＴＥＰ２燃料（濃縮度４．８％）を、各々、取り出し燃焼度３０ＧＷｄ／ｔ、４０ＧＷｄ／ｔ、５０ＧＷｄ／ｔで取り出し回収した計６種の回収ウランを、また、ＢＷＲからは、８×８燃料（濃縮度３．６％）を、取出し燃焼度３０ＧＷｄ／ｔ、４０ＧＷｄ／ｔ、５０ＧＷｄ／ｔで取り出し回収した計３種、及び９×９燃料（濃縮度４％）を、取出し燃焼度３５ＧＷｄ／ｔ、４５ＧＷｄ／ｔ、５５ＧＷｄ／ｔで取り出し回収した計３種の回収ウラン、合計１２種の回収ウランについて行った。これら１２種の回収ウラン及び天然ウランを用いて、ＣＡＮＤＵ炉用の燃料集合体を製造し、運転した。 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. In the evaluation, 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. And 9 × 9 fuel (concentration: 4%) were extracted and recovered at a removal burn-up of 35 GWd / t, 45 GWd / t, and 55 GWd / t for a total of 12 types of recovered uranium. Using these 12 kinds of recovered uranium and natural uranium, a fuel assembly for a CANDU furnace was manufactured and operated.
　表５に、これらの回収ウランの燃料としての価値を示す。ただし、表５において、取り出し燃焼度欄の（）で囲んだ数値、例えば（３５）は、ＢＷＲの９×９燃料の取り出し燃焼度である。 Table 5 shows the value of these recovered uranium as fuel. However, in Table 5, 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.
Figure JPOXMLDOC01-appb-T000005
Figure JPOXMLDOC01-appb-T000005
　図４に、ＰＷＲで燃やされた前記６種の燃料から回収した合計６種の回収ウランと、天然ウランとについて、これらをＣＡＮＤＵ炉で燃やした場合の燃焼度と無限増倍率の変化を概念的に（おおよその様子を）示す。図５に、ＢＷＲで燃やされた前記６種の燃料から回収した合計６種のウランと、天然ウランとについて、これらをＣＡＮＤＵ炉で燃やした場合の燃焼度と無限増倍率の変化を概念的に示す。図４、図５とも、横軸は燃焼度（ＭＷｄ／ｔ）であり、縦軸は無限増倍率（ｋ－ｉｎｆｉｎｉｔｙ）である。 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).
　図４の符号１で示す線は、ＳＴＥＰ２燃料で取り出し燃焼度が３０ＧＷｄ／ｔの場合であり、符号２で示す線はＳＴＥＰ２燃料で取り出し燃焼度が４０ＧＷｄ／ｔの場合とＳＴＥＰ１燃料で取り出し燃焼度が３０ＧＷｄ／ｔの場合であり、符号３で示す線はＳＴＥＰ２燃料で取出し燃焼度が５０ＧＷｄ／ｔの場合とＳＴＥＰ１燃料で取出し燃焼度が４０ＧＷｄ／ｔの場合であり、符号４で示す線はＳＴＥＰ１燃料で取り出し燃焼度が５０ＧＷｄ／ｔの場合と天然ウランの場合である。 The line indicated by reference numeral 1 in FIG. 4 is for STEP2 fuel when the burnup degree is 30 GWd / t, and the line indicated by reference numeral 2 is when STEP2 fuel is taken out and the burnup degree is 40 GWd / t, and 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, and 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.
　また、図５の符号１で示す線は、８×８燃料で取り出し燃焼度が３０ＧＷｄ／ｔの場合と９×９燃料で取出し燃焼度が３５ＧＷｄ／ｔの場合であり、符号２で示す線は８×８燃料で取出し燃焼度が４０ＧＷｄ／ｔの場合と９×９燃料で取出し燃焼度が４５ＧＷｄ／ｔの場合と天然ウランの場合であり、符号３で示す線は８×８燃料で取出し燃焼度が５０ＧＷｄ／ｔの場合と９×９燃料で取出し燃焼度が５５ＧＷｄ／ｔの場合である。 Further, 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, and 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.
　表５及び図４より、ＰＷＲの使用済み燃料からの回収ウランは、おおよそ天然ウランよりも価値が大きいことが判る。また、表５と図６より、ＢＷＲの使用済み燃料からの回収ウランは、到達燃焼度の制限－１０ＧＷｄ／ｔ（８×８燃料であれば４０ＧＷｄ／ｔ、９×９燃料であれば４５ＧＷｄ／ｔ）程度以下の燃焼度で取出されれば、天然ウラン燃料と同等の価値であることが判る。 From Table 5 and FIG. 4, it can be seen that the uranium recovered from the spent fuel of PWR has a greater value than the natural uranium. Further, from Table 5 and FIG. 6, 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.
（実施例２）
　本実施例は、軽水炉から取り出された使用済み燃料を、黒鉛炉に用いることに関する。マグノックス炉や改良型ガス冷却炉においては、減速材として軽水に比較して減速比が大きい黒鉛を用いるため、重水減速炉と同じく天然ウラン燃料を用いている。従って、この場合にも、回収ウラン燃料を用いることが可能となる。ただし、重水と黒鉛の減速比や物理的状態の相違による原子炉の構造等の相違のため、ケースによっては、中性子スペクトル、さらには一回の核***により発生する中性子の個数、発生から消滅までを通じての中性子の核***断面積と捕獲断面積等が多少相違してくる。この結果、回収ウラン燃料の評価に使用する価値にも重水炉を対象とした価値と多少の相違があり得る。 (Example 2)
The present embodiment relates to using spent fuel taken out from a light water reactor for a graphite furnace. In Magnox furnaces and improved gas-cooled furnaces, 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. However, 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.
　しかし、基本は前記ＣＡＮＤＵ炉の場合と同じであり、さらに黒鉛炉において天然ウラン燃料に換えて回収ウランを燃やす場合の検討に必要なデータは、例えば前記日本原子力開発機構の核データライブラリＪＥＮＤＬ－３．３等にて全て公開されている。また、各種の黒鉛炉用に、天然ウラン燃料の燃焼の進行に伴う燃料の無限増倍率の変化を計算する等の各種のプログラムは軽水炉用のものと基本は同じであり、既に開発され、用いられている。あるいは、インプットする数値やデータこそ相違するものの、軽水炉用のプログラムを流用することも可能である。このため、それらのプログラムに天然ウラン燃料の数値（データ）に換えて回収ウラン燃料の数値をインプットすることで、必要な計算を行うことが可能である。そこで、特に特定の黒鉛減速炉を例に採って、具体的な数値を挙げ、図表で説明することは省略する。 However, the basics are the same as in the case of the CANDU furnace, and the data necessary for the examination of burning recovered uranium instead of natural uranium fuel in a graphite furnace is, for example, the nuclear data library JENDL-3 of the Japan Atomic Energy Agency. .3 etc. are all open to the public. For various graphite furnaces, various programs such as calculating the change in infinite multiplication factor of fuel with the progress of combustion of natural uranium fuel are basically the same as those for light water reactors, and have already been developed and used. It has been. Or, although the numerical values and data to be input are different, it is possible to divert the program for light water reactors. For this reason, it is possible to perform necessary calculations by inputting the value of the recovered uranium fuel instead of the value (data) of the natural uranium fuel into these programs. Therefore, taking a specific graphite moderator as an example, a specific numerical value is given and explanation with a chart is omitted.
（実施例３）
　本実施例は、軽水炉で使用済みとされた燃料から回収されたプルトニウムとウランを分離せず、重水炉等で再度燃やすことに関する。 (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.
　前記実施例２までの記載は、軽水炉で使用済みとされた燃料から回収されたウランのみを再度重水炉等で燃やすものであったが、前記燃料には核***性物質であるプルトニウム（主にＰｕ２３９）が１％程度含まれている。このため、軽水炉、例えばＢＷＲで使用済みとされた燃料から核***で生じた娘核種を取り去り、残った（回収した）ウランとプルトニウムを分離せず重水炉の燃料に使用するものである。従って、本実施例の燃料は、回収ウランのみを用いた燃料に比較して、その価値がより大きくなり、ひいては微濃縮ウランを使用するタイプの炉でも問題なく使用可能となる。 In the description up to the second embodiment, only uranium recovered from the fuel used in the light water reactor is burned again in the heavy water reactor or the like, but the fuel contains plutonium (mainly Pu239) which is a fissile material. ) Is included about 1%. For this reason, daughter nuclides generated by nuclear fission are removed from fuel used in light water reactors, for example, BWR, and the remaining (recovered) uranium and plutonium are not separated and used as fuel for heavy water reactors. Therefore, 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.
　ただし、実施例１と比較した場合には、ウランとプルトニムでは、核***により発生する中性子の個数（プルトニウムの方がウランより多い）、核***断面積、中性子の捕獲断面積等の核的性質が相違し、また炉内の中性子スペクトルにも相違が生じてくる。 However, when compared with Example 1, 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.
　しかし、基本は実施例１と同じであり、さらに回収ウラン燃料に換えて回収ウラン－プルトニウム混合燃料を燃やす場合の検討に必要なデータは、具体的には各プルトニム同位体の核種の核的性質は、高速増殖炉の開発に必要なこともあり多くの研究機関、書籍等にて、例えば前記日本原子力開発機構の核データライブラリＪＥＮＤＬ－３．３等にて、全て公開されている。また、重水炉や黒鉛炉を対象として回収ウラン－プルトニウム燃料の燃焼の進行に伴う燃料の無限増倍率の変化を計算する等の各種のプログラムは既に開発されている。即ち、炉心の形状等インプットする事項、数値に多少の相違はあるが、炉心をメッシュ分割して各メッシュ内で単位時間における核反応や中性子束を計算で求め、さらに各メッシュ相互で計算値の遣り取りを行い、次の単位時間で遣り取りを行って修正した値を用いて同じ計算処理を行うことを繰返す等の基本的な計算プロセスは軽水炉と同じであり、このため本実施例に必要な計算を行うことは容易に為し得る。 However, 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. In other words, although there are some differences in input items and numerical values such as the shape of the core, 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.
　さらに、ＢＷＲにおいては、実際にプルトニウムがウランと共に燃やされている。即ち、現在のウラン燃料を燃やすＢＷＲにおいても、各運転サイクルの終わりでは、実際の核***はＵ２３５よりＵ２３８が中性子を捕獲して生じたＰｕ２３９の占める割合が、Ｕ２３５が占める割合よりも多い。従って、ＣＡＮＤＵ炉等で回収ウラン－プルトニウム燃料を燃やすことに特に安全上の問題はない。このため、実施例３と同じく、ＣＡＮＤＵ炉や微濃縮ウランを燃やす炉、例えばあるタイプの黒鉛炉を対象にして具体的な数値を挙げ、図表で説明することは省略する。 Furthermore, in BWR, plutonium is actually burned with uranium. That is, even in the current BWR that burns uranium fuel, at the end of each operation cycle, the actual fission is caused by the fact that Pu239 is generated by U238 capturing neutrons rather than U235. Therefore, there is no particular safety problem in burning the recovered uranium-plutonium fuel in a CANDU furnace or the like. For this reason, as in Example 3, specific numerical values are given for a CANDU furnace and a furnace for burning finely enriched uranium, for example, a certain type of graphite furnace, and description thereof with a chart is omitted.
　実施例１にて示した様に、ＢＷＲはＰＷＲに比べて炉心が大きく、各サイクル運転において、その開始時に炉内における燃料を配置する際の自由度が高い。このため、ＢＷＲから使用済みとして取り出される燃料は、その到達燃焼度の最高値（許容最高燃焼度）に近い状態であることが多い。この結果、ＢＷＲの使用済み燃料から回収されたウランのみを用いた燃料は、天然ウランを用いた燃料に比較して価値が低くなることが多い。しかし、回収ウラン－プルトニウムを用いた燃料であれば、天然ウラン以上の価値が生じることとなるだけでなく、天然ウランを燃やすＣＡＮＤＵ炉はもとより、Ｕ２３５が１～２％の微濃縮ウランを使用する炉においても充分使用可能となる。 As shown in the first embodiment, 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.
（実施例４）
　本実施例は、前記実施例３の変形例であり、解体された核爆弾から取出されたプルトニウムやウランを回収ウラン－プルトニウムに配合して燃やすことに関する。 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.
　前記した様に、ＢＷＲからの回収ウラン－プルトニウムは、ＰＷＲからの回収ウラン－プルトニウムに比較して、その価値は一般的に低い。しかし、核爆弾に用いられるＰｕ２３９やＵ２３５は、９０％以上の濃度を有している。このため、廃棄された核爆弾から取出されたＰｕ２３９やＵ２３５をＢＷＲからの回収ウラン－プルトニウムに僅かに配合した場合、その燃料の価値は大きく上昇し、重水炉や黒鉛炉で燃やすのに最適な価値を有する燃料となる。そして、核爆弾の有効利用、平和的な処理を図ることができる。 As described above, uranium-plutonium recovered from BWR is generally less valuable than uranium-plutonium recovered from PWR. However, 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.
（実施例５）
　本実施例も、前記実施例３の変形例であり、ＢＷＲからの回収ウラン－プルトニウム中のプルトニウムのみに着目してＣＡＮＤＵ炉等で燃やすことに関する。即ち、回収ウラン中のＵ２３５と異なり、回収プルトニウム中のＰｕ２３９、即ち実際に核***をするプルトニウム同位体の濃度はもともと高い。このため、回収されたプルトニウムからＰｕ２３９を濃縮する必要性は全くない。そこで、例えばＣＡＮＤＵ炉で回収プルトニウムを燃やす場合に、天然ウランに配合、あるいは回収ウラン－プルトニウムに配合する等して燃やすことにより、取出し燃焼度の一層の向上を図ることができる。 (Example 5)
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.
　なお、この場合には回収ウランが余るが、余った回収ウランは、高速増殖炉に装荷する、一旦貯蔵施設に保管しておく等の処理を行う。 In this case, recovered uranium remains, but the recovered uranium is loaded into a fast breeder reactor and temporarily stored in a storage facility.

Claims (12)

  1. 　重水炉または黒鉛炉に用いられる原子炉用燃料であって、
    　軽水炉で使用済みとされた燃料から回収された燃料性物質を、前記燃料性物質中に含まれているＵ２３５を濃縮することなく用いて製造されていることを特徴とする原子炉用燃料。     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 a spent fuel in a light water reactor without concentrating U235 contained in the fuel substance.
  2. 　前記軽水炉で使用済みとされた燃料から回収された燃料性物質が、ウランであることを特徴とする請求項１に記載の原子炉用燃料。 The fuel for a nuclear reactor according to claim 1, wherein the fuel substance recovered from the spent fuel in the light water reactor is uranium.
  3. 　前記軽水炉が、ＰＷＲであることを特徴とする請求項１または請求項２に記載の原子炉用燃料。 The nuclear fuel according to claim 1 or 2, wherein the light water reactor is a PWR.
  4. 　前記原子炉用燃料が、ＣＡＮＤＵ炉用の燃料であることを特徴とする請求項１または請求項２に記載の原子炉用燃料。 3. The nuclear reactor fuel according to claim 1 or 2, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
  5. 　前記原子炉用燃料が、ＣＡＮＤＵ炉用の燃料であることを特徴とする請求項３に記載の原子炉用燃料。 4. The nuclear reactor fuel according to claim 3, wherein the nuclear reactor fuel is a fuel for a CANDU reactor.
  6. 　重水炉または黒鉛炉に用いられる原子炉用燃料の製造方法であって、
    　軽水炉で使用済みとされた燃料から燃料性物質を回収する燃料性物質回収ステップと、
    　前記回収された燃料性物質を、前記燃料性物質に含まれているＵ２３５を濃縮することなく用いて燃料を製造する燃料製造ステップと
    を有していることを特徴とする原子炉用燃料の製造方法。     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. Method.
  7. 　前記燃料性物質回収ステップの後に、
    　前記回収された燃料性物質中の各組成物質の組成比を求める組成比算出ステップと、
    　前記各組成物質毎に、核***断面積と核***で生じる中性子の個数との積を捕獲断面積で割り、その商を以て各組成物質の価値係数とする価値係数算出ステップと、
    　前記各組成物質毎に、前記組成比と前記価値係数の積を求め、その総和を以て前記回収された燃料性物質の価値とする価値算出ステップと、
    　前記燃料性物質の価値と所定の基準値とを比較し、その大小に応じて、前記燃料製造ステップへの移行を判断する再利用判断ステップと
    を有していることを特徴とする請求項６に記載の原子炉用燃料の製造方法。     After the fuel substance recovery step,
    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 manufacturing method of the fuel for reactors as described in any one of.
  8. 　重水炉または黒鉛炉に用いられる原子炉用燃料の製造方法であって、
    　軽水炉で使用済みとされた燃料からウランを回収するウラン回収ステップと、
    　前記回収されたウランを、その組成物として含まれているＵ２３５を濃縮することなく用いて燃料を製造する燃料製造ステップと
    を有していることを特徴とする原子炉用燃料の製造方法。     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.
  9. 　前記ウラン回収ステップの後に、
    　前記回収されたウラン中の各ウラン同位体の組成比を求める組成比表算出ステップと、
    　前記各ウラン同位体毎に、核***断面積と核***で生じる中性子の個数の積を捕獲断面積で割り、その商を以て各ウラン同位体の価値係数とする価値係数算出ステップと、
    　前記各ウラン同位体毎に、前記組成比と前記価値係数の積を求め、その総和を以て前記回収されたウランの価値とする価値算出ステップと、
    　前記回収ウランの価値と所定の基準値とを比較し、その大小に応じて、前記燃料製造ステップへの移行を判断する再利用判断ステップと
    を有していることを特徴とする請求項８に記載の原子炉用燃料の製造方法。     After the uranium recovery step,
    A composition ratio table calculating step for obtaining a composition ratio of each uranium isotope in the recovered uranium;
    For each of the uranium isotopes, the product of the fission cross section and the number of neutrons generated by fission is divided by the capture cross section, and the value coefficient calculating step for setting the quotient as the value coefficient of each uranium isotope,
    For each of the uranium isotopes, obtain a product of the composition ratio and the value coefficient, and calculate the value of the value of the recovered uranium with the sum of the products,
    9. 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. The manufacturing method of the fuel for reactors of description.
  10. 　前記軽水炉はＰＷＲであることを特徴とする請求項６ないし請求項９のいずれか１項に記載の原子炉用燃料の製造方法。 The method for producing a nuclear fuel according to any one of claims 6 to 9, wherein the light water reactor is a PWR.
  11. 　前記重水炉または黒鉛炉とはＣＡＮＤＵ炉であることを特徴とする請求項６ないし請求項９のいずれか１項に記載の原子炉用燃料の製造方法。 The method for producing a fuel for a nuclear reactor according to any one of claims 6 to 9, wherein the heavy water reactor or the graphite furnace is a CANDU furnace.
  12. 　前記重水炉または黒鉛炉とはＣＡＮＤＵ炉であることを特徴とする請求項１０に記載の原子炉用燃料の製造方法。 The method for producing a fuel for a nuclear reactor according to claim 10, wherein the heavy water reactor or the graphite furnace is a CANDU furnace.
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Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2017015722A (en) * 2010-02-22 2017-01-19 アドバンスト・リアクター・コンセプツ・エルエルシー Small, fast neutron spectrum nuclear power plant with a long refueling interval

Families Citing this family (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
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

Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH04198891A (en) * 1990-11-29 1992-07-20 Hitachi Ltd Manufacture of fuel assembly, and fuel assembly
JPH07301687A (en) * 1994-03-21 1995-11-14 General Electric Co <Ge> Coating pipe
JPH08170992A (en) * 1994-07-13 1996-07-02 General Electric Co <Ge> Coated pipe and manufacture of coated pipe
JPH09325195A (en) * 1996-06-04 1997-12-16 Power Reactor & Nuclear Fuel Dev Corp Production method for fuel assembly
JP2000292575A (en) * 1999-04-08 2000-10-20 Hitachi Ltd Fuel assembly

Patent Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH04198891A (en) * 1990-11-29 1992-07-20 Hitachi Ltd Manufacture of fuel assembly, and fuel assembly
JPH07301687A (en) * 1994-03-21 1995-11-14 General Electric Co <Ge> Coating pipe
JPH08170992A (en) * 1994-07-13 1996-07-02 General Electric Co <Ge> Coated pipe and manufacture of coated pipe
JPH09325195A (en) * 1996-06-04 1997-12-16 Power Reactor & Nuclear Fuel Dev Corp Production method for fuel assembly
JP2000292575A (en) * 1999-04-08 2000-10-20 Hitachi Ltd Fuel assembly

Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2017015722A (en) * 2010-02-22 2017-01-19 アドバンスト・リアクター・コンセプツ・エルエルシー Small, fast neutron spectrum nuclear power plant with a long refueling interval

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