JP2018124123A - Method for reducing degree of harmfulness of spent fuel, mox fuel aggregate, and mox fuel reactor core - Google Patents

Abstract

[PROBLEMS] To reduce the degree of harmfulness without equipment specialized for separation and conversion of specific nuclides and nuclide conversion. A method for reducing the harmfulness of spent fuel includes a step S01 for extracting spent uranium fuel after burning in a light water reactor, a step S02 for deriving a harmfulness reduction condition for the spent uranium fuel, and a reduction condition If it is determined that the reduction condition is satisfied, step S04 for reprocessing spent uranium fuel and recovering uranium, plutonium and MA, and nuclear fuel material recovered in the reprocessing step And MA and newly added nuclear fuel material, step S05 for producing MOX fuel, and step S06 for taking out the used MOX fuel after burning in the light water reactor, and repeating the MOX fuel combustion step after the judgment step . The total concentration of fissile nuclides in the MOX fuel is within a predetermined range determined for each MOX fuel combustion step for the spent MOX fuel. [Selection] Figure 1

Description

  Embodiments described herein relate generally to a method for reducing the harmfulness of spent fuel, a MOX fuel assembly, and a MOX fuel core.

  Spent nuclear fuel generated during the operation of the reactor contains high levels of radioactive waste and has a high radiotoxicity over a long period of time. The value converted to the amount of exposure when it is assumed that the nuclide contained in the spent nuclear fuel is orally ingested is called the potential radiotoxicity or potential hazard and is used as an indicator of the hazard of waste. The potential radiotoxicity of spent nuclear fuel is caused by fission products (FP) and transuranium elements (TRU) contained in the fuel. Among these nuclides, long-term The influence of TRU becomes dominant.

  Since the storage and disposal of radioactive waste requires a great deal of time and money, it is required to reduce the potential radiotoxicity of these spent nuclear fuels.

  For example, as a means of reducing the potential radiotoxicity of TRU, there is a method of converting TRU into a short half-life nuclide. As a method for converting TRU into a short half-life nuclide, a method of performing nuclear transmutation using fast neutrons in a fast reactor is known. However, at present, fast reactors for transmutation have not been put into practical use and mass production, and a long development period is expected to be realized. For this reason, it is difficult to reduce the potential radiotoxicity contained in spent fuel generated from currently operated light water reactors using a fast reactor.

  In the case of a once-through method in which spent nuclear fuel is not reprocessed, all the nuclides contained in the spent nuclear fuel are treated as waste. On the other hand, in the pluthermal method, plutonium and uranium are recovered, reprocessed and used as a mixed oxide (MOX) fuel, and reused in a light water reactor. In a general pull thermal process, TRUs other than plutonium, that is, minor actinides (MA) are separated from plutonium and uranium that are reused. Thus, in a light water reactor, transmutation of MA once taken out is not assumed.

When performing nuclear transmutation in a light water reactor, increase the transmutation efficiency of nuclides that contribute significantly to potential radiotoxicity such as Pu ²³⁸ (plutonium 238), Pu ²⁴¹ (plutonium 241), Am ²⁴¹ (Americium 241), etc. There is a need. However, in general light water reactor fuels and light water reactor cores, fuel design and core design have not been performed from the viewpoint of efficiently performing MA transmutation.

  Generally, in a light water reactor fuel and a light water reactor core, the fuel is designed and the reactor is operated so that the excess reactivity is zero at the end of one operating cycle, that is, EOC (End of Cycle). .

Japanese Patent No. 5743518 Japanese Patent No. 2804205 Japanese Patent No. 3960572

Nuclear Handbook 1st edition 3rd edition, Ohmsha Light Water Reactor Fuel Behavior Nuclear Safety Research Association (edition), 4th edition Separation and conversion technology general first edition 1st edition, Japan Atomic Energy Society

  As described above, uranium isotopes and plutonium isotopes are extracted from spent fuel used in the core of the light water reactor in the process of the pull thermal method, and FP and MA are discarded as radioactive waste.

  As a method for reducing the potential radiotoxicity in these radioactive wastes, a method of separating a harmful nuclide and converting the nuclide to a less harmful nuclide, that is, separation conversion is known.

  Typical methods assumed for separation conversion are harmful by adding harmful nuclides to MOX fuel and burning them in a fast reactor, or by using an accelerator to target harmful nuclides. A method of converting a new nuclide into a less harmful nuclide.

  The above-described method requires advanced reprocessing technology to separate specific nuclides, and it is necessary to construct a dedicated fast reactor and accelerator for the conversion of nuclides. The development of these technologies and the construction of equipment has been a problem that requires enormous time and cost.

  Embodiments of the present invention have been made to solve the above-described problems, and do not perform separation conversion of specific nuclides, and reduce harmfulness without developing equipment specialized for nuclide conversion. The purpose is to do.

  In order to achieve the above-mentioned object, an embodiment of the present invention is a method for reducing the harmfulness of spent fuel, which reduces the harmfulness caused by transuranium elements generated from spent fuel in a light water reactor, and burns uranium fuel in a light water reactor. Then, a uranium fuel combustion step for taking out the used uranium fuel, a reduction condition derivation step for deriving a harmfulness reduction condition for the spent uranium fuel, and a harmfulness reduction condition for the spent uranium fuel. It is determined whether or not the condition is satisfied, and if it is determined that the harmfulness reduction condition is not satisfied, the determination step of moving to disposal is performed. If it is determined that the harmfulness reduction condition is satisfied, the spent uranium fuel is reused. A reprocessing step for recovering and recovering uranium, plutonium and minor actinides, and the recovered uranium and recovered plutonium recovered in the reprocessing step. And a MOX fuel production step for producing MOX fuel from the recovered minor actinide and newly added nuclear fuel material, and a MOX fuel combustion step for taking out the used MOX fuel after burning the MOX fuel in a light water reactor. Then, the determination step is repeated until the MOX fuel combustion step, and the total concentration of fissile plutonium and fissile uranium in the MOX fuel is a predetermined value determined for each MOX fuel combustion step for the spent MOX fuel. It is in the range.

  An MOX fuel assembly according to an embodiment of the present invention includes a plurality of MOX fuel rods extending in the longitudinal direction and containing nuclear fuel material containing fissile plutonium and fissile uranium and arranged in parallel to each other, and in the longitudinal direction. And a plurality of fuel rods containing flammable poisons containing uranium or plutonium as main components and a combustible poison and arranged in a grid with the MOX fuel rods, and having a predetermined burnup The initial value of the total concentration of the fissile plutonium and the fissile uranium is set so that the surplus reactivity at the end of the secured operation cycle is greater than a predetermined positive value.

  An MOX fuel core according to an embodiment of the present invention includes a plurality of the MOX fuel assemblies arranged in a grid in parallel with each other, and a plurality of controls respectively arranged in the array of the plurality of MOX fuel assemblies. And a bar.

  According to the embodiments of the present invention, it is possible to reduce the degree of harmfulness without performing separation and conversion of specific nuclides and without developing equipment specialized for nuclide conversion.

It is a flowchart which shows the whole procedure of the harmfulness reduction method of the spent fuel which concerns on 1st Embodiment. It is a graph which shows the example which calculated the change of the potential radiotoxicity after 100 years with respect to the change of the initial Puf enrichment as an effect of the harmfulness reduction method of the spent fuel which concerns on 1st Embodiment. It is a graph which shows the change of the normalized radiotoxicity normalized about the example shown in FIG. It is a graph which shows the relative value of the result shown in FIG. It is a graph explaining the predetermined conditions which ensure the effect of the harmfulness reduction method of the spent fuel which concerns on 1st Embodiment. It is a conceptual explanatory view of a spent fuel harmfulness reduction possible range in the spent fuel harmfulness reduction method according to the first embodiment. It is a horizontal sectional view showing the configuration of the MOX fuel assembly according to the first embodiment. It is a horizontal sectional view showing the configuration of the MOX fuel core according to the first embodiment. It is a flowchart which shows the whole procedure of the harmfulness reduction method of the spent fuel which concerns on 2nd Embodiment. It is a horizontal sectional view showing the composition of the uranium fuel core used in the method for reducing the degree of harmfulness of spent fuel according to the second embodiment. It is a horizontal sectional view showing the composition of the enrichment increase uranium fuel assembly using enrichment enrichment uranium fuel used in the spent fuel harmfulness reduction method according to the second embodiment. It is a graph which shows the relative value of the change of the potential radiotoxicity after 100 years with respect to the change of the enrichment of the initial stage ^U235 which shows the effect of the harmfulness reduction method of the spent fuel which concerns on 2nd Embodiment. It is a graph which shows the change of the infinite multiplication factor with respect to the change of the burnup explaining the harmfulness reduction method of the spent fuel which concerns on 2nd Embodiment.

  Hereinafter, a method for reducing the degree of harmfulness of spent fuel, an MOX fuel assembly, and an MOX fuel core according to an embodiment of the present invention will be described with reference to the drawings. Here, the same or similar parts are denoted by common reference numerals, and redundant description is omitted.

[First Embodiment]
FIG. 1 is a flowchart showing the entire procedure of the spent fuel harmfulness reduction method according to the first embodiment.

First, conventional uranium fuel is loaded into a light water reactor and burned in the light water reactor, and then taken out of the reactor (step S01). Here, the conventional uranium fuel is, for example, uranium dioxide (UO ₂ ) fuel in which the enrichment of U ²³⁵ (uranium 235) is about 3 to 5%. Moreover, the combustion of uranium fuel has an average burnup of, for example, about 40 to 50 GWd / t. A higher burnup may also be used.

  Next, after the combustion in the light water reactor in step S01, the spent fuel harmfulness reduction condition is derived for the used uranium fuel (used U fuel) taken out from the light water reactor (step S02). Here, the harmfulness reduction condition of the spent fuel is a condition for reducing the harmfulness of the spent fuel. This will be specifically described in the description with reference to FIG.

  Next, it is determined whether the spent U fuel taken out from the light water reactor used in step S01 satisfies the harmfulness reduction condition of the spent fuel derived in step S02 (step S03).

  If it is determined that the spent U fuel does not satisfy the harmfulness reduction condition of the spent fuel (NO in step S03), the spent U fuel is not reprocessed and the process proceeds to disposal (step S07). To do.

  If it is determined that the spent fuel harmfulness reduction condition is satisfied for the spent U fuel (YES in step S03), the process proceeds to the next reprocessing step (step S04).

  In the reprocessing step of step S04, the spent uranium fuel taken out from the core of the light water reactor is cooled for an appropriate period, and then reprocessed. In the reprocessing, MA is recovered together with uranium and plutonium containing fissile nuclides.

  Next, recovered minor actinide (recovered MA), which is recovered MA, is added to recovered uranium (recovered U), which is recovered uranium, and recovered plutonium (recovered Pu), which is recovered plutonium. Plutonium (Pu), natural uranium (natural U) or enriched uranium (enriched U) is added to produce an MOX fuel to which MA is added, that is, a mixed oxide fuel of Pu and U (step S05). In the reprocessing, MA may be recovered together with uranium or plutonium without being separated from uranium or plutonium.

  Here, the total amount of U and Pu in the initial fissionable nuclides of MOX fuel, namely, U fissionable nuclides (fissile uranium: fissile U) and Pu fissile nuclides (fissile plutonium: fissile Pu). The ratio to the total amount, that is, the value of the total concentration is a value greater than or equal to a value necessary for ensuring a predetermined burnup. In the following description, it is assumed that the newly added nuclear fuel material is Pu. For this reason, the enrichment degree of Pu to be newly added is a value necessary for securing the necessary amount of fissile nuclide.

  Next, after the MOX fuel assembly to which MA has been added is loaded on the MOX fuel core of the light water reactor and burned, the spent MOX fuel is taken out of the MOX fuel core (step S06).

  Next, returning to step S03, it is determined whether or not the spent MOX fuel taken out from the MOX fuel core satisfies the harmfulness reduction condition of the spent fuel, and is determined to satisfy the harmfulness reduction condition (step S03). Only in the case of YES), the above steps S03 to S06 are repeated. Of the above steps, steps S04 to S06 are referred to as a recycling step. The number of recycles is counted so that the first pass through these steps and the second pass through these steps.

  FIG. 2 is a graph showing an example of a trial calculation of a change in potential radiotoxicity after 100 years with respect to a change in the initial Puf enrichment as an effect of the method for reducing the harmfulness of spent fuel according to the first embodiment. . That is, the potential radiotoxicity after 100 years of MA in spent marrow fuel is used as an example of an index representing the harmfulness of spent fuel.

The horizontal axis in FIG. 2 represents the initial Puf enrichment (wt%), that is, the Puf enrichment of the MOX fuel before combustion in the furnace at each recycle number. Here, the Puf enrichment is the amount of Pu fissile nuclides (Fissile Pu) such as Pu239 (Pu ²³⁹ ) with respect to the total weight of the nuclear fuel material mainly containing Pu and U in the MOX fuel. It is a ratio of weight. In this trial calculation example, the newly added nuclear fuel material is Pu, and the existence of fissile nuclides includes U235 in the recovered uranium, in addition to the recovered Pu and the newly added Pu fissionable nuclides. There are fissile nuclides.

  The vertical axis in FIG. 2 is the potential radiotoxicity (Sv / MWd) after 100 years for MA in spent MOX fuel taken after irradiation in the MOX fuel core at the number of recycles.

  The four broken lines in FIG. 2 correspond to the number of recycling times of 1 to 4 times. That is, the solid line connecting point A1, point B1, and point C1 corresponds to the number of times of recycling, and the dotted line connecting point A2, point B2, and point C2 corresponds to the number of times of recycling, and points A3, B3, and A broken line connecting the points C3 corresponds to the number of times of recycling, and a one-dot chain line connecting the points A4, B4, and C4 corresponds to the number of times of recycling of 4.

  Here, as described above, the number of recycles refers to the number of times that the MA generated by the combustion of the first uranium fuel has undergone reprocessing and recovery thereafter. That is, the point in time when the noted MA is collected by reprocessing spent U fuel is the first recycling. After that, when the spent MOX fuel is first reprocessed and collected, the second time is the number of recycling times. Next, when the spent MOX fuel is used and irradiated and then reprocessed and collected, the third time is the number of recycling times. , And so on.

  The points on each broken line are points indicating conditions and results. That is, for each point on the broken line, the value on the horizontal axis is the condition of the initial Puf enrichment, and the value on the vertical axis is the result of irradiating the MOX fuel to a predetermined burnup and used MOX The potential radiotoxicity after 100 years for MA in fuel is shown. Hereinafter, specific explanation will be given by taking several points as examples.

  Point A1 indicates the case where the number of times of recycling is 1, that is, the case of the first recycling. The value of the initial Puf enrichment on the horizontal axis at the point A1 indicates the Puf enrichment at the stage where the MOX fuel is manufactured by reprocessing spent uranium fuel. The value of the vertical axis at point A1 indicates the potential radiotoxicity after 100 years of MA in the spent MOX fuel taken out from the furnace after the MOX fuel is burned in the furnace and a predetermined burnup is obtained. Indicates the value of. That is, it is a value when the process is shifted to the disposal process without being irradiated in the furnace using the new MOX fuel.

  The value of the initial Puf enrichment is an enrichment such that the MOX fuel can obtain a predetermined burnup. In this trial calculation example, the predetermined burnup is 45 GWd / t, which is often used in normal design, and the initial Puf enrichment value for securing this burnup is 6.0 wt% in this trial calculation example. It is. For this purpose, Pu is newly added to the recovered U, recovered Pu, recovered MA, and the like after the reprocessing of the spent U fuel, thereby securing an initial Puf enrichment of 6.0 wt%.

  As a result of irradiation in the MOX fuel core with a burnup of 45 GWd / t, the potential radiotoxicity after 100 years of MA in the used spent MOX fuel is about 10,000 Sv / MWd.

  Next, point A2 indicates the case of the second recycling. That is, this is a case where the MA taken out after the first recycling is used again for the production of MOX fuel without shifting to the disposal process.

  In this case, in order to ensure the burnup of 45 GWd / t, in this trial calculation example, the initial Puf enrichment needs to be about 8 wt%. For this purpose, Pu is newly added to produce an MOX fuel with an initial Puf enrichment of about 8 wt%.

  In the second MOX fuel for recycling, Pu is newly added as described above, thereby generating new MA. In addition, there is a decrease due to the original MA irradiation. As a result, the potential radiotoxicity after 100 years of MA in the extracted spent MOX fuel is about 19,000 Sv / MWd.

  Similarly, the point A3 indicates the case of the third recycling, and the initial Puf enrichment for securing the burnup of 45 GWd / t is about 10 wt%. Also, as a result of irradiation in the MOX fuel core with a burnup of 45 GWd / t, the potential radiotoxicity after 100 years of MA in the extracted used MOX fuel is about 27,000 Sv / MWd.

  Similarly, for point A4, the initial Puf enrichment is about 11 wt%, and the potential radiotoxicity after 100 years of MA in the used spent MOX fuel is about 35,000 Sv / MWd.

  As described above, at point A1, MOX fuel having an initial Puf enrichment of 6.0 wt%, which is an initial Puf enrichment for securing a burnup of 45 GWd / t, is set as an initial condition, and then each recycle is performed. In order to obtain an initial Puf enrichment for obtaining a burnup of 45 GWd / t and sequentially irradiating the MA with neutrons in the furnace, the points A2, A3, and The condition and result indicated by point A4 are obtained. This flow is called an A series.

  On the other hand, each of points B1 and C1 is a case where the initial condition is that the initial Puf enrichment exceeding the initial Puf enrichment for securing the burnup of 45 GWd / t is set. That is, at the point B1, the initial Puf enrichment is about 13.4 wt%. At point C1, the initial Puf enrichment is about 22.4 wt%.

  By repeating the same recycling as the A series with the initial Puf enrichment of the point B1 being about 13.4 wt%, the B series of points B2, B3, and B4 is obtained. Further, when the same recycling as the A series is repeated with the condition that the initial Puf enrichment of the point C1 is about 22.4 wt% as the first condition, a C series of points C2, C3, and C4 is obtained.

  In each of A series, B series, and C series, as the number of recycling shifts from 1 to 4, the initial Puf enrichment in the case of the next recycling count compared to the previous recycling count, respectively. The degree and the potential radiotoxicity after 100 years will increase sequentially.

  Now, let's compare the case of not recycling multiple times and the case of recycling multiple times. In the case of multiple recycles, the potential radiotoxicity after 100 years of MA recovered from spent MOX fuel after the last recycle will remain.

  On the other hand, when not being recycled a plurality of times, that is, when the number of times of recycling is one, MA in the used MOX fuel is generated every time power is generated. Therefore, when power generation is performed a plurality of times, the MA in the used MOX fuel is generated each time the number of times of recycling is one.

  Therefore, based on the result of dividing the potential radiotoxicity after 100 years of MA recovered from spent MOX fuel after the last number of recycles in the case of multiple recycles by the number of recycles, we will compare each other. It is considered appropriate to do so. Now, this result will be referred to as potential radiotoxicity 100 years after standardization (hereinafter, “normalized radiotoxicity”).

  FIG. 3 is a graph showing changes in normalized radiotoxicity normalized for the example shown in FIG.

  Here, each point on the polygonal lines L1, L2, L3, and L4 in FIG. 3 is a point indicating a condition and a result as in FIG. Specifically, the points on each broken line are the same as the points shown in FIG. 2 in that the horizontal axis indicates the condition of the initial Puf enrichment, but the vertical axis indicates the predetermined combustion of the MOX fuel. 2 shows the normalized radiotoxicity for MA in spent MOX fuel, which is the result after irradiation to a degree. For each broken line, the solid line connecting points a1, b1 and c1 has a recycling count of 1, the dotted line connecting points a2, b2 and c2 has a recycling count of 2, and the broken line connecting points a3, b3 and c3 has a recycling count of 1 3 times and the dashed-dotted line which connects the points a4, b4, and c4 show the case where the frequency | count of recycle is 4 times.

  Therefore, point a1, point a2, point a3, point a4, point b1, point b2, point b3, point b4, point c1, point c2, point c3 and point c4 in FIG. 3 are point A1, point A2 in FIG. , Point A3, point A4, point B1, point B2, point B3, point B4, point C1, point C2, point C3 and point C4, respectively.

  As shown in FIG. 3, the normalized radiotoxicity in each series decreases as the number of recycling increases in the A series, B series, and C series.

  FIG. 4 is a graph showing the relative values of the results shown in FIG. The horizontal axis represents the initial Puf enrichment (wt%). The vertical axis is the relative value of potential radiotoxicity after 100 years of normalization. Specifically, in each of the A series, the B series, and the C series, the number of times of recycling is 1, that is, a value based on the potential radiotoxicity after irradiation in the first MOX fuel core.

  As shown in FIG. 4, the trend of change in relative value of potential radiotoxicity after 100 years normalized relative to the increase in initial Puf enrichment at each number of recycles changes as the number of recycles increases. ing.

  That is, when the number of recycles is 1, the increase in the initial Puf enrichment, that is, the normalized radiotoxicity increases from the point a1 to the point b1, and slightly decreases at the point c1. When the number of recycles is 2, the normalized radiotoxicity increases as the initial Puf enrichment increases, that is, from point a2 to point b2, and decreases when it reaches point c2.

  In addition, when the number of recycles is 3, the increase in the initial Puf enrichment, that is, the normalized radiotoxicity increases slightly from the point a3 to the point b3, and decreases at the point c3 from the case of the point c1. . Further, when the number of recycles is 4, the normalized radiotoxicity is not substantially changed even when the initial Puf enrichment is increased, that is, from the point a4 to the point b4, and decreases when the point c4 is reached.

  As described above, the normalized radiotoxicity tends to decrease as the initial Puf enrichment increases as the number of recycling increases. In particular, in the region where the value of the initial Puf enrichment is large, the degree of decrease in normalized radiotoxicity with respect to the increase in the number of recyclings is significant.

  That is, as shown in FIG. 3, the normalized radiotoxicity decreases as the number of recycling increases, and as the initial Puf enrichment increases as shown in FIG. 4, the normalized radiotoxicity decreases.

  However, as shown in FIG. 3, at the stage where the number of recycles is small, the increase in the normalized radiotoxicity with respect to the increase in the initial Puf enrichment is large, so that the normalized radiotoxicity is higher than the standardized radiotoxicity after the first recycle. In order to decrease, it is necessary to have a predetermined condition of a combination of at least a predetermined number of recycling times and a predetermined initial Puf enrichment for the first MOX fuel, or a condition exceeding the predetermined condition.

  FIG. 5 is a graph for explaining predetermined conditions for ensuring the effect of the method for reducing the harmfulness of spent fuel.

  Now, for point a1, the value on the horizontal axis is the Puf enrichment required for the MOX fuel produced after reprocessing spent uranium fuel and recovering U, Pu and MA to ensure a predetermined burnup Degree. The value on the vertical axis for point a1 is the normalized radiotoxicity of MA when the MOX fuel is burned in the MOX fuel core, the spent MOX fuel is reprocessed, recovered and transferred to the disposal side. is there. Based on this condition and result, in order to say that a certain state has the effect of reducing the harmfulness, in FIG. 5, the normalized radiotoxicity, which is the value on the vertical axis at the point representing the final condition and result, It is necessary to be smaller than the normalized radiotoxicity Rna1, which is the value on the vertical axis at the point a1.

  In FIG. 5, in the case of the broken line L4 when the number of recycles is 4, the vertical Puff enrichment region including the points a4, b4, and c4 from the point a4 to the point c4 The normalized radiotoxicity that is the value of the axis is smaller than the normalized radiotoxicity Rna1 at the point a1. Therefore, when the number of recycles is 4, if the initial Puf enrichment of the MOX fuel after the reprocessing of spent uranium fuel, which is the first condition, is equal to or higher than the initial Puf enrichment PEa1 at the point a1, There will be a reduction in harmfulness.

  Next, in FIG. 5, in the case of the broken line L3 when the number of recycles is 3, the value of the vertical axis is smaller than the value of the vertical axis at the point a1 because the initial Puf rich from the point a4 to the point c4. Of the degree of conversion, only a part of the region where the initial Puf enrichment is large. That is, since the point d3 is a point where the normalized radiotoxicity value on the vertical axis is equal to the point a1, there is an effect of reducing the harmfulness in the region from the point d3 to the point c3 without including the point d3. Similar to the curve Lb connecting the B series points and the curve Lc connecting the C series points, if the curve Ld connecting the D series points giving the point d3 is considered, the first condition and the point representing the result are d1. Is obtained as shown in FIG. Therefore, when the number of recycles is 3, if the initial Puf enrichment level of the MOX fuel after the reprocessing of the spent uranium fuel, which is the first condition, exceeds the initial Puf enrichment level PEd1 at the point d1 This will have an effect of reducing harmfulness.

  FIG. 6 is a conceptual explanatory diagram of a possible range for reducing the harmfulness of spent fuel in the method for reducing the degree of harmfulness of spent fuel according to the present embodiment. The horizontal axis in FIG. 6 represents the initial fissionable nuclide ratio in each recycling. If the added fissile element is Pu, the initial fissionable nuclide ratio is, for example, the initial Puf enrichment, and this case will be described below. The vertical axis represents the number of recyclings.

  Regarding the initial Puf enrichment, the Puf enrichment necessary for obtaining a normal predetermined burnup is indicated by x0. On the other hand, if the initial Puf enrichment is too high, the margin is reduced with respect to the thermal limit condition and the nuclear limit condition of the core, and further, the feasibility cannot be ensured. Therefore, the initial Puf enrichment has a feasible upper limit value Xm.

  The position shown in FIG. 6 where the number of recycles is the largest and the initial Puf enrichment is the largest is the limit condition Q. Let Xm0 be the initial Puf enrichment in the first number of recyclings of the limit series corresponding to this condition. The maximum limit condition of the initial Puf enrichment that is smaller than or equal to the polygonal line connecting these conditions is expressed as a maximum limit condition for the ratio of the fissionable nuclide ratio. Here, generalization means that the newly added fissionable nuclide may be not only fissile Pu but also fissile U, or the sum of both.

  On the other hand, there is a minimum initial Puf enrichment to ensure a predetermined burnup during the recycle period. The minimum initial Puf enrichment condition for each number of recycles is displayed as a generalized fissile nuclide ratio minimum limit condition for the minimum limit condition of the initial Puf enrichment that requires a greater enrichment than the polygonal line connecting each recycling number. ing.

  When the recycling is repeated, the total amount of MA increases as shown in FIG. 2, and there is no solution that can be established as MOX fuel. For this reason, there is an upper limit to the number of times of recycling. FIG. 6 shows the recyclable frequency condition.

  In the example of FIG. 6, the region satisfying the above restriction conditions is a region surrounded by a thick line with one to six recycles. In this, the range where the normalized radiotoxicity after 100 years after standardization is smaller than the potential radiotoxicity Rna1 after 100 years under the reference conditions is indicated by a shaded area. The range of this region is the range in which the potential radiotoxicity can be reduced, that is, the range in which the harmfulness of the spent fuel can be reduced.

  The range in which the harmfulness of the spent fuel can be reduced depends on conditions such as the composition of the uranium fuel and the combustion history in the light water reactor in step S01, the subsequent composition of the MOX fuel, the combustion history in the MOX fuel core, and the like. For this reason, the harmfulness reduction condition of the spent fuel derived in step S02 may be reviewed at any time if there is a change in the combustion plan in the MOX fuel core.

  FIG. 7 is a horizontal sectional view showing the configuration of the MOX fuel assembly according to the first embodiment. The MOX fuel assembly 10 includes a plurality of MOX fuel rods 11 containing a nuclear fuel material containing fissile plutonium and fissile uranium, a plurality of fuel rods 12 containing a combustible poison containing a combustible poison, and two water rods 13. And a channel box 15 on the outside in the radial direction for storing them.

  The plurality of MOX fuel rods 11 and the plurality of combustible poison-containing fuel rods 12 extend in the longitudinal direction and are arranged in parallel to each other to form a 10 × 10 grid-like array. The two water rods 13 are arranged at the center of the lattice-like arrangement of the MOX fuel rods 11 and the fuel rods 12 with combustible poisons.

  Each of the plurality of MOX fuel rods 11 includes a MOX fuel element 11a and a cladding tube 11b that accommodates the MOX fuel element 11a.

  The MOX fuel element 11a is made of, for example, a plurality of pellets that are cylindrical and stacked in the axial direction. In each pellet of the MOX fuel element 11a, MOX fuel obtained by adding nuclear fuel material to the recovered U, recovered Pu, and recovered MA recovered by reprocessing the used U fuel or the used MOX fuel is sintered in a columnar shape. Has been molded.

  Pu or U is used as the nuclear fuel material. The total concentration of fissionable nuclides in recovered U, recovered Pu and newly added nuclear fuel material, i.e. fissile Pu and fissile U, i.e. the concentration of fissile nuclides with respect to the total weight of Pu and U, is determined in the MOX fuel core. It is a concentration higher than a concentration at which a predetermined burnup, for example, a burnup of 45 GWd / t in the above-described trial calculation example can be secured by combustion.

  Further, the concentration of the fissile nuclide is a concentration that can reduce the harmfulness after a plurality of recyclings. Specifically, in the trial calculation example shown in FIG. 5, the concentration of the fissile nuclide for the first MOX fuel after the reprocessing of the spent U fuel is higher than about 6 wt% in terms of Puf enrichment. .

  Each of the plurality of flammable poison-containing fuel rods 12 includes a flammable poison-containing fuel element 12a and a cladding tube 12b for housing the fuel element 12a. In the combustible poisoned fuel element 12a, the total concentration of fissile nuclides is lower than the total concentration of fissile nuclides in the MOX fuel element 11a.

  As described above, regarding the configuration of the MOX fuel assembly 10, the MOX fuel rods 11 and the fuel rods 12 containing flammable poisons are arranged in a 10 × 10 lattice pattern and two water rods 13 are shown. It is not limited. For example, an 8 × 8 or 9 × 9 lattice array or a triangular array may be used. For example, the number of water rods 13 may be one.

  FIG. 8 is a horizontal sectional view showing the configuration of the MOX fuel core according to the first embodiment. The MOX fuel core 1 is arranged in the center of a plurality of MOX fuel assemblies 10 arranged in a grid and forming a substantially cylindrical shape as a whole, and four MOX fuel assemblies 10 in a 2 × 2 arrangement adjacent to each other. A plurality of control rods 5 are provided.

  In FIG. 8, the core in which the MOX fuel assembly 10 is loaded shows the case where all the fuel assemblies are the MOX fuel assemblies 10, but the present invention is not limited to this. For example, the MOX fuel assembly 10 and the core loaded with a normal uranium fuel assembly may be burned.

  As described above, according to the present embodiment, the degree of harmfulness can be reduced without performing separation conversion of specific nuclides and without developing equipment specialized for nuclide conversion.

[Second Embodiment]
FIG. 9 is a flowchart showing the entire procedure of the spent fuel harmfulness reduction method according to the second embodiment. This embodiment is a modification of the first embodiment.

  In step S01 in the present embodiment, instead of loading the conventional uranium fuel in the first embodiment into the light water reactor and burning it in the light water reactor, the enriched uranium fuel having a higher enrichment than the conventional uranium fuel is used. To be loaded and burned. The enrichment-enhanced uranium fuel is enriched so that the core has a surplus reactivity exceeding a predetermined positive value even at the end of the light water reactor operation cycle. Details will be described below.

  FIG. 10 is a horizontal sectional view showing the configuration of the uranium fuel core used in the method for reducing the degree of harmfulness of spent fuel according to the second embodiment.

  The uranium fuel core 2 is arranged in the center of a plurality of enriched uranium fuel assemblies 20 that are arranged in a grid and form a substantially cylindrical shape as a whole, and four enriched uranium fuel assemblies 20 adjacent to each other. It has a plurality of control rods 5 arranged.

  In FIG. 10, the core in which the enriched uranium fuel assembly 20 is loaded shows a case where all the fuel assemblies are enriched uranium fuel assemblies 20, but the present invention is not limited to this. For example, irradiation may be performed with a core in which the MOX fuel assembly is partially loaded together with the enriched uranium fuel assembly 20.

  FIG. 11 is a horizontal sectional view showing a configuration of an enriched uranium fuel assembly using enriched uranium fuel used in the method for reducing the degree of harmfulness of spent fuel according to the second embodiment.

  The enrichment enriched uranium fuel assembly 20 includes a plurality of enrichment enriched uranium fuel rods 21, a plurality of combustible poison-containing fuel rods 12, two water rods 13, and radially outside of these to accommodate them. A channel box 15 is provided.

  The plurality of enriched uranium fuel rods 21 and the plurality of combustible poison-containing fuel rods 12 are arranged in parallel to each other to form a grid-like arrangement. The water rod 13 is arranged at the center of the grid-like arrangement of the enriched uranium fuel rods 21 and the burnable poison-containing fuel rods 12.

  Each of the plurality of enriched uranium fuel rods 21 includes an enriched uranium fuel element 21a and a cladding tube 21b that accommodates the enriched uranium fuel element 21a.

  The enrichment increasing uranium fuel element 21a is made of, for example, a plurality of pellets that are cylindrical and stacked in the axial direction. The enriched uranium fuel element 21a uses enriched uranium, and the enrichment has an enrichment necessary to ensure a predetermined burnup, for example, a value greater than 3.8 wt%.

  As described above, with respect to the configuration of the enriched uranium fuel assembly 20, the enriched uranium fuel rod 21 and the combustible poison-containing fuel rod 12 are arranged in a 10 × 10 grid and the number of the water rods 13 is two. Although shown, it is not limited to this. For example, an 8 × 8 or 9 × 9 lattice array or a triangular array may be used. For example, the number of water rods 13 may be one.

  FIG. 12 is a graph showing the relative value of the change in the potential radiotoxicity after 100 years with respect to the change in the enrichment of the initial U235 showing the effect of the method for reducing the harmfulness of the spent fuel according to the second embodiment. The horizontal axis is the initial U235 enrichment, that is, the enrichment of U235 in the U fuel at the start of combustion in the core. The vertical axis represents the relative value of potential radiotoxicity 100 years after the completion of combustion in the core. The relative value is a relative value based on 3.8 wt% as a general U fuel enrichment. As shown in FIG. 12, the higher the initial U235 enrichment, the lower the potential radiotoxicity.

  FIG. 13 is a graph illustrating a change in infinite multiplication factor with respect to a change in burnup, which explains the method for reducing the harmfulness of spent fuel according to the second embodiment. The horizontal axis represents the burnup of uranium fuel and enriched uranium fuel. The burn-up to be achieved by the uranium fuel assembly and the enrichment-enhanced uranium fuel assembly 20 is 45 GWd / t in terms of the average burn-out. Until this burnup is reached, the light water reactor has passed three operation cycles from the first cycle to the third cycle, and at the end of the third run cycle, a burnup of a predetermined burnup, in this case 45 GWd / t. Achieve the degree.

  In both the uranium fuel assembly and the enriched uranium fuel assembly 20, the flammable poison is added at a concentration that burns in the first cycle.

  The broken line shown by the solid line is 3.8 wt% where the initial enrichment of U235 is generally used, and the broken line shown by the broken line increases the enrichment by 5.0 wt% of the initial enrichment of U235. Show the case.

  In this case, in the first cycle of operation, the infinite multiplication factor increases as the burnup increases due to combustion of the flammable poison. In the second and third cycles, the infinite multiplication factor decreases due to the burning of fissile nuclides in the fuel. When the initial enrichment of U235 is 3.8 wt%, the infinite multiplication factor becomes almost zero with the end of the third cycle. On the other hand, the infinite multiplication factor when the initial enrichment of U235 is 5.0 wt% is larger than the infinite multiplication factor when the initial enrichment is 3.8 wt% at the end of the third cycle. Δk∞ is about 0.07.

  From the viewpoint of excess reactivity, the initial enrichment level of U235 is about 0.3% Δk or less for the excess reactivity to be secured at the end of the operation cycle and the analysis of the excess reactivity by combustion analysis etc. In the design analysis, it is effective that the excess reactivity remains at least about 0.3% Δk or more. Therefore, it is effective to set the predetermined value to a value larger than 0.3% Δk, for example, 2% Δk.

  As described above, when the operation cycle length and the extraction average burnup are constant, the higher the initial U235 enrichment, the higher the concentration of U235 contained in the extracted fuel at the end of the operation cycle. Therefore, the concentration of U235 is higher in the recovery U recovered by the reprocessing step of step S04 than in the case of the first embodiment. For this reason, in the production of MOX fuel that can obtain the necessary burnup in the next recycling, the amount of fissile nuclides to be replenished can be reduced. As a result, in manufacturing the MOX fuel, the enrichment of plutonium can be reduced as compared with the case of the first embodiment.

[Other Embodiments]
As mentioned above, although embodiment of this invention was described, embodiment is shown as an example and is not intending limiting the range of invention.

  Furthermore, these embodiments can be implemented in various other forms, and various omissions, replacements, and changes can be made without departing from the scope of the invention. These embodiments and their modifications are included in the scope and gist of the invention, and are also included in the invention described in the claims and the equivalents thereof.

  DESCRIPTION OF SYMBOLS 1 ... MOX fuel core, 2 ... Uranium fuel core, 5 ... Control rod, 10 ... MOX fuel assembly, 11 ... MOX fuel rod, 11a ... MOX fuel element, 11b ... Cladding tube, 12 ... Fuel rod with combustible poison, 12a ... Fuel element with combustible poison, 12b ... cladding tube, 13 ... water rod, 15 ... channel box, 20 ... enriched uranium fuel assembly, 21 ... enriched uranium fuel rod, 21a ... enriched uranium fuel Element, 21b ... cladding tube

Claims (6)

A method for reducing the harmfulness of spent fuel to reduce the harmfulness caused by transuranium elements generated from spent fuel in a light water reactor,
A uranium fuel combustion step in which uranium fuel is burned in a light water reactor and then used uranium fuel is taken out;
A reduction condition derivation step for deriving a reduction degree of harmfulness for the spent uranium fuel;
For the spent uranium fuel, it is determined whether or not a harmfulness reduction condition is satisfied, and if it is determined that the harmfulness reduction condition is not satisfied, a determination step of shifting to disposal;
A reprocessing step of reprocessing the spent uranium fuel and recovering uranium, plutonium and minor actinides, if determined to satisfy the hazard reduction conditions;
A MOX fuel production step for producing MOX fuel from the recovered uranium, recovered plutonium and recovered minor actinide recovered in the reprocessing step, and a newly added nuclear fuel material;
A MOX fuel combustion step of taking out the used MOX fuel after burning the MOX fuel in a light water reactor;
Have
After the determination step, the MOX fuel combustion step is repeated.
The total concentration of fissile plutonium and fissile uranium in the MOX fuel is within a predetermined range determined for each MOX fuel combustion step for the spent MOX fuel.
A method for reducing the degree of harmfulness of spent fuel.   The harm reduction conditions are greater than the fissile nuclide ratio minimum limit condition value in each recycling step and less than the fissile nuclide ratio maximum limit condition value, and the number of recycles is less than or equal to the recyclable upper limit number of times. The method for reducing the harmfulness of spent fuel according to claim 1, comprising:   The enrichment of the uranium fuel burned in the uranium fuel combustion step is a value such that the surplus reactivity at the end of the operation cycle is greater than a predetermined positive value. The method for reducing the harmfulness of spent fuel as described in 1. A plurality of MOX fuel rods extending in the longitudinal direction and containing nuclear fuel material containing fissile plutonium and fissile uranium and arranged in parallel with each other;
A plurality of fuel rods containing a flammable poison, which includes a nuclear fuel material and a flammable poison that extend in the longitudinal direction and mainly contains uranium or plutonium and are arranged in a grid with the MOX fuel rod;
Have
The initial value of the total concentration of the fissile plutonium and fissile uranium is set so that the excess reactivity at the end of the operation cycle in which the predetermined burnup is ensured is greater than a predetermined positive value. Characteristic MOX fuel assembly.   The MOX fuel assembly according to claim 4, wherein an initial value of the total concentration is greater than 6 wt%. A plurality of MOX fuel assemblies according to any one of claims 4 and 5 arranged in a grid in parallel with each other;
A plurality of control rods each disposed in an array of the plurality of MOX fuel assemblies;
A MOX fuel core comprising:

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