JP4417859B2 - Transport container for radioactive material - Google Patents

Description

  The present invention relates to a transport container for radioactive materials such as spent nuclear fuel.

  As described in Non-Patent Document 1, the transport storage container for radioactive materials such as spent nuclear fuel from nuclear power plants, etc. is a wet type container in which internal water is enclosed, and a dry type that does not enclose internal water in the container Although both are roughly classified into containers, in order to reduce radiation exposure to the environment and the exposure of workers engaged in transport operations, the radiation dose rate is 2 millisieverts per hour (mSv on the surface of the package). / H), the gamma rays and neutrons emitted from the radioactive material are shielded so as to satisfy the limit of 0.1 millisievert per hour (mSv / h) at a point of 1 meter from the surface. At the same time, it is configured to efficiently dissipate heat generated when radioactive materials such as spent nuclear fuel stored inside collapse.

  That is, the wet container shields gamma rays with a container body made of iron or lead, and neutrons are formed by internal water sealed inside the container body and a resin-based neutron shielding material (resin) disposed around the container body. And heat is dissipated by the internal water and heat transfer fins that penetrate the neutron shielding material. As described in Patent Document 1, the dry container is a container in which a gas such as helium is filled in the container body instead of internal water, and a neutron shielding material is a resin-based neutron shielding material disposed around the container body. The heat is dissipated by heat transfer fins that penetrate the neutron shielding material.

N. Aoki, "All about radioactive material transport", 2nd edition, Nikkan Kogyo Shimbun, April 30, 2002, p. 137-140 JP 2002-250790 A

  Although the resin-based neutron shielding material described above shields neutrons, it has a low thermal conductivity, and therefore inhibits heat dissipation from the container body. Therefore, as described above, the transport storage container is configured such that heat dissipation is promoted by the heat transfer fins that penetrate the neutron shielding material. That is, in order to achieve both the neutron shielding function and the heat dissipation function, the structure of the transport storage container is complicated, which increases the manufacturing cost.

  An object of the present invention is to provide a transport container for a radioactive material with a simple structure and low manufacturing cost without impairing the neutron shielding function.

Means and effects for solving the problems

The radioactive substance transport container of the present invention has a function of shielding gamma rays, and has a container body having an internal space for containing the radioactive substance immersed in internal water, and decelerates neutrons emitted from the radioactive substance. is formed of an inorganic material which, after slowing the neutrons so as to advance in the internal water, have a, an inner shield which is disposed around the radioactive material, said inner shield is the radioactive substance A basket having a plurality of grids containing a fuel assembly including the inner assembly, wherein the inner shield is configured by connecting a plurality of divided bodies arranged in contact with the outer surface of the basket along the outer surface. The container body has a cylindrical shape, and the basket housed in the internal shield is placed when the container body containing the internal shield is placed down. The number of the grid located at the top of bets, characterized in that more than the number of the grid located at the bottom.

  According to the present invention, since the internal shield actively decelerates the neutrons released from the radioactive material, the neutrons can be efficiently shielded with the internal water, and the resin for shielding neutrons as in the past. It is not necessary to provide a neutron shielding material for the system around the container body. This eliminates the need for providing heat transfer fins for promoting heat dissipation through the neutron shielding material having low thermal conductivity, and heat generated from the fuel assembly is efficiently radiated from the container body. Therefore, the structure of the radioactive material transport container can be simplified without impairing the neutron shielding function, the manufacturing cost can be reduced, and disassembly can be facilitated.

In addition , by connecting a plurality of segments that can be easily placed in close contact with the outer surface of the basket along the outer surface, it is possible to easily produce an internal shield that is in close contact with the outer surface of the basket, and neutrons are divided first. Since it progresses to internal water after penetrating, the neutron moderating effect by the internal shield can be effectively obtained.

In addition , when the container body is tilted down during transportation or the like, a space for relaxing expansion of internal water and an increase in internal pressure in the container body extends in the horizontal direction above the internal space. Since the state where the internal shield is immersed in the internal water can be maintained, the neutron shielding function by the internal water is not impaired on the upper side of the internal space.

  In the present invention, the inorganic material forming the internal shield may be any metal of iron, iron-based alloy, copper, and copper-based alloy. According to this, it is possible to form an internal shielding body having an excellent neutron moderating effect with a general-purpose material and at a low cost. Moreover, since these materials also have a gamma ray shielding function, they are more suitable as an internal shield.

  In the present invention, the internal shield may have a thickness of 40 mm to 120 mm. According to this, it is possible to suppress an increase in the burden of attaching the internal shield to the basket while preferably obtaining a neutron moderating effect.

  Hereinafter, embodiments of a radioactive substance transport container according to the present invention will be described with reference to the drawings. FIG. 1 is a cross-sectional view of a radioactive material transport container. 2 is a cross-sectional view taken along the line AA in FIG. FIG. 3 is an explanatory diagram of the internal shield. FIG. 4 is a graph showing a simulation result of a dose equivalent rate of 1 m from the container surface when the thickness of the internal shield is changed.

  A container body 2 of a radioactive material transport container (hereinafter simply referred to as a transport container) 1 has a bottomed cylindrical shape having an internal space 8 and is made of carbon steel or stainless steel having a gamma ray shielding function. The container body 2 has a minimum thickness necessary for functioning as a sealed container having a gamma ray shielding function. By setting the required minimum thickness in this way, the weight of the entire transport container 1 is reduced while sufficiently exhibiting the gamma ray shielding function. Further, heat generated from a fuel assembly described later is efficiently radiated from the outer surface of the container body 2.

  One end 2 a side of the container body 2 is sealed with a lid 9 made of the same material as the container body 2. Note that a gasket (not shown) is provided between the container body 2 and the lid 9 in order to ensure the sealing performance of the transport container 1 as a pressure vessel.

  On the one end 2a side of the container main body 2, two upper trunnions 6 project in the vertical direction and two in the horizontal direction (three shown in the figure) with an interval of 90 degrees in the circumferential direction. ing. The upper trunnion 6 is used to suspend the transport container 1 with the one end 2a side up and the other end 2b side down. Thus, the transportation container 1 can be suspended in a stable state by suspending the transportation container 1 at four positions spaced by 90 degrees.

  On the other end 2 b side of the container body 2, two lower trunnions 7 (one shown in the figure) project in the horizontal direction with an interval of 180 degrees in the circumferential direction. The lower trunnion 7 is used to suspend the transport container 1 with the upper end 2c on the upper side and the lower end 2d on the lower side. At this time, the two upper trunnions 6 protruding in the horizontal direction are simultaneously used.

  A water level adjusting vent orifice 3 is provided on the upper end 2c side of the container body 2 on the one end 2a side. A drain orifice 4 is provided on the upper end 2c side of the container body 2 and on the other end 2b side. Further, the lid 9 is provided with a convex portion 9a and a pressurizing valve 5 for suspending the lid 9.

  The water level adjusting vent orifice 3 is configured such that the transport container 1 is erected with the one end 2 a side up and the other end 2 b side down, the lid 9 is removed, and the internal water 11 is injected into the internal space 8 of the container body 2. Opened. Of the internal water 11 accumulated from the bottom b of the internal space 8 due to the opening of the water level adjusting vent orifice 3, the portion exceeding the water level a is discharged from the water level adjusting vent orifice 3. The water level does not exceed the water level a. Thereby, if the container 9 is covered with the lid 9, a space by gas (air) is secured between the water level a and the bottom of the lid 9. Then, as shown in FIG. 2, when the sealed transport container 1 is in a transported state (side-down state), a gas (air) space 12 extends in the horizontal direction above the internal space 8. Such a space 12 is used to mitigate expansion of the internal water 11 and an increase in pressure in the internal space 8 due to heat generation of the fuel assembly. Needless to say, the transport container 1 lies horizontally without being affected by the upper trunnion 6 protruding from the lower end 2d side of the container body 2 when the container 1 is put down.

  In FIG. 1, the position where the water level adjusting vent orifice 3 is provided, that is, the water level a can secure a space 12 that can alleviate the expansion of the internal water 11 and the pressure increase in the internal space 8, and Thus, when the transport container 1 is placed down, the inner shield 13 described later is sufficiently adjusted so as to be immersed in the internal water 11.

  On the other hand, in FIG. 1, the drain orifice 4 is provided at the position of the bottom b of the internal space 8, with the transport container 1 standing with the one end 2 a side up and the other end 2 b side down, Opened when the internal water 11 is discharged to the outside of the container body 2. Due to the opening of the drain orifice 4, all the internal water 11 in the internal space 8 is discharged from the drain orifice 4.

  In FIG. 2, an internal shield 13 that accommodates a basket 10 having 15 long grids (square pipes) in close contact is accommodated in the internal space 8 of the container body 2. Further, internal water 11 having a neutron shielding function is sealed in the internal space 8. A space 12 for relieving expansion of the internal water 11 and an increase in pressure in the internal space 8 is secured above the internal space 8. The number of lattices that the basket 10 has is not limited to this.

  The basket 10 is made of boron-added stainless steel, and a rod-shaped fuel assembly (not shown) is accommodated in each lattice. In the example of FIG. 2, the lattice arrangement of the basket 10 is such that the uppermost number of lattices is larger than the lowermost number of lattices. Specifically, four lattices are provided at the top, five at the bottom of the top, four at the bottom, and two at the bottom. By making the basket 10 an asymmetrical shape with a lower height on the upper side with respect to the center P of the internal space 8 of the transport container 1, the internal space when the transport container 1 is placed down during transport or the like. A space 12 made of gas (air) extends in the horizontal direction above 8, but the state in which the internal shield 13 containing the basket 10 in close contact with the internal water 11 can be maintained. Therefore, the neutron shielding function by the internal water 11 is not impaired on the upper side of the internal space 8.

  The internal shield 13 that accommodates such a basket 10 is formed of an inorganic material having an effect of decelerating neutrons emitted from the fuel assembly. Since the internal shield 13 actively decelerates neutrons emitted from the fuel assembly accommodated in the basket 10 at an early stage, the internal water 11 can efficiently shield the neutrons. The reason why the internal shield 13 is made of an inorganic material is that an organic material is deteriorated by radiation even if it contains a lot of hydrogen that decelerates neutrons. Moreover, among inorganic materials, it is desirable that the metal material has an excellent neutron moderating effect. This is because such a metal material has a gamma ray shielding function at the same time.

That is, the inorganic material forming the internal shield 13 is preferably any metal of iron, iron-based alloy, copper, and copper-based alloy. This is because it is a general-purpose material at low cost and can form the internal shield 13 having an excellent neutron moderating effect, and since these materials also have a gamma ray shielding function, it is more suitable as the internal shield 13. is there. In particular, iron has a relatively high density of 7.85 g / cm ³ and is most often used as a shielding material for gamma rays, and has a higher neutron moderating effect than lead, which is also frequently used as a shielding material for gamma rays, and copper. Less expensive. Lead not only is inferior in neutron moderation effect compared to iron, but also contaminates the internal water 11, and is therefore not preferable as the internal shield 13 in terms of environmental pollution. Considering the above, iron and iron-based alloys are more preferable as the inorganic material forming the internal shield 13. In the present embodiment, the internal shield 13 is made of stainless steel, but may be carbon steel.

  3A is a front view of the internal shield 13, FIG. 3B is a top view of the internal shield 13, FIG. 3C is an XX cross-sectional view of FIG. 3A, and FIG. ) Is a bottom view of the internal shield 13, and FIG. 3E is a rear view of the internal shield 13. As shown in FIGS. 3A to 3D, the internal shield 13 is composed of 12 plate-like divided bodies 13a to 13l centered on the basket 10 (not shown). While being arranged on the outer side surface, it is arranged along the outer side surface, and the adjacent ones are connected by bolts. In this way, by connecting twelve divided bodies 13a to 13l that can be easily arranged on the outer surface of the basket 10 along the outer surface, the inner shield 13 that contacts the outer surface of the basket 10 can be easily manufactured. Since neutrons first pass through the divided bodies 13a to 13l and then proceed to the internal water 11, the neutron moderating effect by the internal shield 13 can be effectively obtained. In addition, since intensity | strength is not requested | required of the internal shielding body 13, from the viewpoint of cost reduction, the division bodies 13a-13l which comprise the internal shielding body 13 may overlap | superpose the some rolling material. Moreover, the division | segmentation number of the division bodies 13a-13l is not limited to this. Further, the shape of the divided body is not limited to a plate shape, and can be selected in consideration of workability and the like such as a rod shape, a block shape, a tile shape, or a combination thereof.

  In addition, the thickness of the divided bodies 13a to 13l, that is, the thickness of the internal shield 13 is preferably 40 mm or greater and 120 mm or less. With this thickness, it is possible to suppress an increase in the burden of attaching the internal shield 13 to the basket 10 while suitably obtaining a neutron moderating effect.

  Further, as shown in FIG. 3 (e), a support that supports the fuel assembly so as not to come out of the basket 10 is provided on the back surface 16 a of the inner shield 13 that becomes the bottom when the transport container 1 is set up. Metal fittings 14a to 14d are provided. Specifically, the support fittings 14a support four fuel assemblies housed in the four uppermost lattices of the basket 10, respectively, and the support fittings 14b are five pieces one lower than the uppermost stage of the basket 10. Each of the five fuel assemblies housed in the lattices of the basket 10 is supported, and the support fitting 14c supports the four fuel assemblies housed in the four lattices, two lower than the uppermost stage of the basket 10, respectively. The support fitting 14d supports the two fuel assemblies housed in the two lowermost lattices of the basket 10, respectively.

  Further, as shown in FIGS. 3A to 3E, the internal shield 13 supports the internal shield 13 in the internal space 8 by contacting the inner peripheral surface of the container body 2. Support bodies 15a to 15e are provided. Specifically, the support 15a is provided on the plate-like divided body 13c, the support 15b is provided on the plate-like divided body 13g, the support 15c is provided on the plate-like divided body 13k, and the support 15d. , 15e are provided on a plate-like divided body 13a. Thereby, when the internal shield 13 is inserted into the internal space 8 from the back surface 16a side, the support bodies 15a to 15e come into contact with the inner peripheral surface of the container body 2, so that the internal shield 13 is within the internal space 8. While being able to prevent moving around, the internal shield 13 can be supported under the surface of the internal water 11 even when the transport container 1 is placed down. Further, as shown in FIGS. 3B to 3D, each of the supports 15a to 15e on the back surface 16a side is provided with a smooth insertion of the internal shield 13 into the internal space 8. A taper is provided.

  Next, the operation of the radioactive material transport container in the above configuration will be described. In a nuclear power plant or the like, the transport container 1 suspended by a wire hooked on the upper trunnion 6 is vertically submerged in the fuel pool, and the internal space 8 is filled with the internal water 11 poured from the fuel pool. A fuel assembly that has been cooled in the fuel pool for a predetermined period is accommodated in each lattice of the basket 10 in the container body 2. Thereafter, the lid 9 is closed and the transport container 1 is lifted out of the fuel pool. Thereafter, in the decontamination pit, the water level of the internal water 11 is adjusted to the water level a by the water level adjusting vent orifice 3, and the transport container 1 is sealed by fixing the lid 9 to the container body 2 with a bolt. Thereafter, the transport container 1 is carried on a loading platform such as a truck, laid down, and transported by a truck or the like.

  At this time, as shown in FIG. 2, a space 12 for relaxing expansion of the internal water 11 and an increase in internal pressure in the container body 2 extends in the horizontal direction above the internal space 8. Since the internal shield 13 to be accommodated is immersed in the internal water 11, neutrons emitted from the fuel assembly are decelerated by the internal shield 13 and are suitably shielded by the internal water 11. Further, the gamma rays emitted from the fuel assembly are shielded by the container body 2, and the heat generated from the fuel assembly is efficiently radiated from the outer surface of the container body 2.

  The transport container 1 according to the present embodiment is not limited to the one used for off-site transportation from a nuclear power plant or the like to a reprocessing facility or the like, and is used for on-site transfer of the same nuclear power plant or the like. Also good.

  Next, in order to confirm the effect of the transport container in the present embodiment, the following test was performed. That is, 15 units of boiling water reactor (BWR) fuel having a enrichment of U-235 (uranium 235) of 3.6% and a burnup of about 36800 MWd / MTU are stored in the basket 10. In a state where the transport container 1 was placed down, a dose equivalent rate of 1 m was simulated from the container surface when cooling was performed for 180 days. Here, the inner shield 13 is made of carbon steel, the total thickness of the carbon steel constituting the container body 2 and the inner shield 13 is maintained at 38 cm, and the thickness of the inner shield 13 and the container body 2 The thickness ratio was changed. Here, the distance c from the upper end of the internal shield 13 to the water surface of the internal water 11 in FIG. 2 is set to 6.5 cm, but the distance c is appropriately optimized according to the required neutron shielding ability. Thus, it is not necessarily 6.5 cm. In addition, the reference value of the dose equivalent rate of 1 m from the container surface in the domestic transport regulations is 100 μSv / h. The result is shown in FIG.

  From FIG. 4, it can be seen that the thickness of the internal shield 13 only needs to be 40 mm or more in order to satisfy the reference value. Moreover, when the thickness of the internal shield 13 is 120 mm or more, the dose equivalent rate reduction effect tends to converge gradually. Therefore, considering that the burden of attaching the internal shield 13 to the basket 10 increases if the internal shield 13 becomes too thick, the thickness of the internal shield 13 is in the range of 40 mm to 120 mm. preferable.

  Next, in the above test conditions, the dose equivalent rate of 1 m from the container surface when the thickness of the inner shield 13 is maintained at 40 mm and the material of the inner shield 13 is changed to aluminum, copper, and carbon steel, respectively. Was simulated. The results are shown in Table 1.

  From Table 1, it can be seen that copper and carbon steel have approximately the same neutron moderation effect and gamma ray shielding function. Therefore, it can be said that iron and iron-based alloys, which are cheaper than copper, are more preferable. On the other hand, it can be seen that aluminum is inferior to carbon steel and copper in both the neutron moderating effect and the gamma ray shielding function.

  As described above, the radioactive substance transport container 1 according to the present embodiment has a function of shielding gamma rays, and has a container body 2 having an internal space 8 in which the radioactive substance is immersed in the internal water 11. The inner shield 13 is formed of an inorganic material that decelerates neutrons emitted from the radioactive substance, and is disposed around the radioactive substance so that the neutrons are decelerated and then proceed to the internal water 11. ing. According to this, since the internal shielding body 13 installed in the position close to the radioactive substance actively decelerates the neutron emitted from the radioactive substance, the internal water 11 can efficiently shield the neutron, Thus, there is no need to provide a resin-based neutron shielding material for shielding neutrons around the container body. This eliminates the need to provide heat transfer fins for promoting heat dissipation through the neutron shielding material having low thermal conductivity, and heat generated from the fuel assembly is efficiently radiated from the outer surface of the container body 2. The Therefore, the configuration of the radioactive material transport container 1 can be simplified without impairing the neutron shielding performance, the manufacturing cost can be reduced, and the disassembly can be facilitated.

  Further, the internal shield 13 contains a basket 10 having a plurality of lattices that contain fuel assemblies containing radioactive substances, and the internal shield 13 is arranged in contact with the outer surface of the basket 10. A plurality of the divided bodies 13a to 13l are arranged along the outer surface. According to this, the inner shield 13 in contact with the outer surface of the basket 10 can be easily manufactured by connecting a plurality of divided bodies 13a to 13l that can be easily arranged on the outer surface of the basket 10 along the outer surface. Since neutrons first pass through the divided bodies 13a to 13l and then proceed to the internal water 11, the neutron moderating effect by the internal shield 13 can be effectively obtained.

  Further, the shape of the internal space 8 of the container body 2 is a columnar shape, and when the container body 2 containing the internal shield 13 is placed down, it is positioned at the uppermost stage of the basket 10 accommodated in the internal shield 13. The number of lattices is larger than the number of lattices located at the lowermost stage. According to this, when the container main body 2 is placed down during transportation or the like, the space 12 for relaxing the expansion of the internal water 11 and the increase in the internal pressure in the container main body 2 is provided above the internal space 8. Although extending in the horizontal direction, the state in which the internal shield 13 is immersed in the internal water 11 can be maintained, so that the neutron shielding function by the internal water 11 is not impaired on the upper side of the internal space 8.

  In addition, the inorganic material forming the internal shield 13 is made of any metal of iron, iron-based alloy, copper, and copper-based alloy. According to this, it is possible to form the internal shield 13 having a good neutron moderating effect with a general-purpose material and at a low cost. Moreover, since these materials also have a gamma ray shielding function, they are more suitable as the internal shield 13.

  Further, the thickness of the internal shield 13 is set to 40 mm or more and 120 mm or less. According to this, it is possible to suppress an increase in the burden of attaching the internal shield 13 to the basket 10 while preferably obtaining a neutron moderating effect.

  Moreover, although this invention was demonstrated based on suitable embodiment, this invention can be changed in the range which does not exceed the meaning. That is, the inner shield may not be configured by connecting a plurality of flat plate-like divided bodies arranged in contact with the outer surface of the basket along the outer surface. In this case, what is necessary is just to comprise an internal shielding body using the L-shaped and U-shaped plate-shaped body separately processed. Further, if neutrons are suitably shielded, the inner shield is not limited to the one in contact with the outer surface of the basket, and may be disposed in close proximity without being in contact with the inner shield, but between the inner shield and the basket. Since it is possible to reduce the size by eliminating the dead space, or to easily secure the thickness of the internal water layer, it is more preferable to dispose the internal shield so as to contact the outer surface of the basket.

  2 is preferable from the viewpoint of suppressing the cost increase, but as shown in FIG. 5, the number of lattices located in the uppermost stage of the basket accommodated in the internal shield is located in the lowermost stage. It may not be more than the number of lattices. FIG. 5 shows an internal shield having 17 lattices, two at the top, four at the bottom of the top, five at the bottom, four at the bottom, and two at the bottom. Is provided. In this case, the two uppermost grids are not immersed in the internal water, and the neutrons released upward are not shielded by the internal water. Therefore, an additional auxiliary shield is provided on the upper outer periphery of the container body. It is necessary to shield the neutron. In addition, by setting the center of the basket accommodated in the internal shield to a position that is eccentric downward with respect to the center of the internal space of the transport container in the transported state, the basket is used as internal water while ensuring sufficient space above. If it is configured to be immersed, the auxiliary shield can be omitted.

  Moreover, the inorganic material which forms an internal shielding body may not be any metal among iron, an iron-type alloy, copper, and a copper-type alloy, for example, may use lead. In this case, although the cost is increased, in order to avoid contamination of the internal water, it is sufficient to use a lead sealed in, for example, stainless steel.

  Further, the thickness of the internal shield may not be 40 mm or more and 120 mm or less. In this case, if the thickness is 40 mm or more, neutrons can be suitably decelerated even if the thickness is 120 mm or more, so that the neutron shielding function by the internal water can be sufficiently exhibited.

It is sectional drawing of the transport container of a radioactive substance. It is AA sectional drawing of FIG. It is explanatory drawing of an internal shielding body, (a) is a front view, (b) is a top view, (c) is XX sectional drawing of (a), (d) is a bottom view, (e) is a rear view. It is. It is a graph showing the simulation result of the dose equivalent rate of 1 m from the container surface when the thickness of the internal shield is changed. It is AA sectional drawing of FIG.

Explanation of symbols

1 Transport container (transport container for radioactive material)
2 container body 2a one end 2b other end 2c upper end 2d lower end 3 water level adjusting vent orifice 4 drain orifice 5 pressurizing valve 6 upper trunnion 7 lower trunnion 8 internal space 9 lid 9a convex portion 10 basket 11 internal water 12 space 13 internal shield 13a-13l Divided bodies 14a-14d Support brackets 15a-15e Supports

Claims (3)

A container body having a function of shielding gamma rays and having an internal space for storing radioactive substances immersed in internal water;
An internal shield that is formed of an inorganic material that decelerates neutrons emitted from the radioactive substance, and is disposed around the radioactive substance so that the neutron is decelerated and then proceeds to the internal water;
I have a,
The inner shield contains a basket having a plurality of grids containing fuel assemblies containing the radioactive material;
The inner shield is configured by connecting a plurality of divided bodies arranged in contact with the outer surface of the basket along the outer surface,
The shape of the internal space of the container body is a columnar shape,
When the container main body containing the internal shield is placed down, the number of the lattices located at the uppermost stage of the basket accommodated in the internal shield is greater than the number of lattices located at the lowermost stage. A transport container for radioactive materials, characterized by many . The radioactive material transport container according to claim 1, wherein the inorganic material forming the internal shield is one of iron, iron-based alloy, copper, and copper-based alloy. The transport container for radioactive substances according to claim 1 or 2 , wherein the thickness of the internal shield is 40 mm or more and 120 mm or less.

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