JPS61124894A - Store of spent fuel and manufacture of said store - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

DETAILED DESCRIPTION OF THE INVENTION [Field of Application of the Invention] The present invention relates to a spent fuel storage body and a method for manufacturing the storage body.

[Background of the invention]

Spent fuel from nuclear power plants goes through storage and reprocessing processes.

Regarding storage, currently it is stored in a pool within the reactor, but due to delays in the reprocessing plan, it has become necessary to temporarily store it for several decades to 100 years.

In this case, high density storage is an essential requirement, whether stored in an in-furnace pool, in a cage or in a separate storage facility. Furthermore, as a method for reprocessing medium-sized spent fuel rods, as shown in Japanese Patent Application Laid-open No. 151395/1983, hydrogen is stored in the zirconium alloy cladding tube of the spent fuel rod, and the cladding tube is crushed to produce nuclear fuel. Indicates separation from matter.

However, the processing capacity of fuel reprocessing equipment is not inferior, and the majority of spent fuel is currently stored within nuclear power plants. Spent fuel can be stored in a pool or in a dry cask. The pool storage method requires a cooling water purification system, which increases the size of the equipment, and there are problems in that it is not easy to carry in and out. The dry cask storage method eliminates the disadvantages of the pool storage method, but it has the problem of low decay heat removal capacity because it is a dry method, and the containers are large and heavy due to neutron and R-ray radiation. I'm holding it.

[Purpose of the invention]

An object of the present invention is to provide a spent fuel storage body with high radiation shielding ability and a method for manufacturing the storage body.

Fill the center and place 1 piece in the container surrounding the filling area.
It consists in filling with zirconium hydride.

[Embodiments of the invention]

A spent fuel storage body, which is a preferred embodiment of the present invention, is
This will be explained based on the diagram and FIG.

The spent fuel storage body 1 is constructed by filling a container 2 with UO, which is a spent nuclear fuel material, pellets 3, zirconium hydride 4, and a radiation shield 5. Nuclear fuel material 3 is extracted from spent fuel rods, and is stored in container 2.
is filled in the center. The zirconium hydride 4 is obtained by absorbing hydrogen into a zirconium alloy cladding tube of a spent fuel rod and pulverizing it, and is filled into the container 2 so as to surround the filling area of the nuclear fuel material 3. There is.

A radiation shield 5ti is filled in the container 2 so as to surround the zirconium hydride 4. The radiation shield 5 includes a substance that blocks r-rays and a substance that blocks neutrons. Since it is filled with zirconium hydride 4, the amount of neutron shielding material is reduced. The neutron shielding material may be a neutron absorbing material.

Since such a spent fuel storage body 1 can utilize a neutron shielding material using hydrogen of zirconium hydride, the thickness of the radiation shielding body 5 can be made thin. Therefore, a larger amount of the constituent material of the spent fuel rod can be filled.

The method for manufacturing the spent fuel storage body in this example will be described below. The apparatus shown in FIG. 3 is for crushing zirconium alloy parts of spent fuel rods. This apparatus consists of a hydrogen treatment container 10 which can be kept airtight with a packing 9. Heater 11, hydrogen gas cylinder 12, exhaust pump 1
3. Nuclear fission product gas processing device 7, hydrogen gas storage tank 1
4, a pressure regulator 15, a powder filter 16, and an exhaust gas trap 17. The hydrogenation treatment vessel 10 includes a support member 28 for holding the spent fuel rods therein, and a receiving plate 27 disposed below the support member 28 and made of wire mesh.
A receiver made of a plate and placed at the lowest level has a plate 29. The support member 28 is composed of a plurality of round bars arranged at intervals. A slide plate 8 that slides horizontally is provided immediately below the support member 28.

A large number of spent fuel rods 7 removed from a spent fuel assembly taken out from the core of a nuclear reactor are placed in a hydrogen treatment vessel 10.
Place it on the support member 28 inside. The spent fuel rod 7 has a zirconium alloy cladding tube filled with a large number of UOs and pellets. The spent fuel assembly itself is supported by the support member 2B.
You can place it on top. Then, the upper lid of the hydrogenation treatment container 10 is closed. At this time, the slide plate 8
is inserted below. The hydrogen gas in the hydrogen gas cylinder 12 is transferred to the hydrogen treatment container 1 without passing through the FP gas treatment device 7.
Introduce to 0. The pressure of hydrogen gas is regulated by a pressure regulator 15. After hydrogen is supplied, when the temperature inside the hydrogenation treatment vessel 10 is raised by the heater 11, hydrogen storage in the zirconium alloy members such as the zirconium alloy cladding that constitutes the spent fuel rod begins. Zirconium alloy members are fractured by hydrogen embrittlement. At this time, as shown in Figure 4, 30
When the temperature is above 0C, hydrogen storage becomes active. According to Fig. 4, the hydrogen storage capacity decreases again on the high temperature side, but the hydrogen diffusion rate in the zirconium alloy member increases, so 30
If it is 0tll' or more, there is no problem.

Thereafter, when the pressure inside the hydrogenation treatment vessel 10 is lowered by the exhaust pump 13, a hydrogen desorption phenomenon occurs as shown in FIG. Unable to withstand the change in lattice constant from zirconium hydride, metallic zirconium turns into powder. Gas discharged from the hydrogen treatment vessel 10 passes through a fission product gas processing device 7 in order to collect the fission product gas discharged from the fuel rods 1 . As a fission product gas processing device,
There are cold traps and traps using various adsorbents. For example, a liquid nitrogen trap can collect gases other than hydrogen and helium. In the case of a cold trap, the fission product gas is collected in the cooler 18 and transferred to the collection device, or the heater 19 is used to transfer it to the exhaust gas trap 17 depending on the situation, such as when the radioactivity has sufficiently attenuated. Can be transported.

Further, hydrogen can be recycled and used after passing through the storage tank 14 and the pressure regulating device 15.

Since the fission gas does not affect the hydrogen absorption/desorption cycle, it is not necessary to completely collect the fission gas. When the above hydrogen absorption/desorption cycle is repeated,
Enables thorough atomization of zirconium alloy members. The powder filter 16 is installed to prevent fine powder from flowing into the exhaust system during gas exhaust.

In addition, the oxide film (Z r Q *) on the surface of the cladding tube, etc.
and cladding attached to the surface (mainly Fe*
Os) is also converted into metal powder by hydrogen ring reduction during the above treatment. UO and pellets are rounded heat generating elements (≧400C) and can reduce the amount of external heating during the hydrogen absorption/desorption process. In particular, after hydrogen embrittlement of the zirconium alloy member by hydrogen absorption, the UOt pellet and the zirconium hydride come into mixed contact, so the amount of external heating by the heater 11 in the subsequent hydrogen desorption process can be significantly reduced.

After repeating hydrogen storage and desorption operations on the zirconium alloy member of the spent fuel rod 1 several times in the hydrogenation treatment vessel 1θ, hydrogen is supplied into the hydrogenation treatment vessel 10 and hydrogen is stored in the zirconium alloy member. The crushing operation is stopped at the stage where the zirconium alloy member is turned into a zirconium hydride. And slide plate 8
When the support member 28 is pulled out, the UOt pellets and the shattered zirconium hydride fall onto the tray 27. Since the receiver plate 27 is made of a wire mesh, the zirconium hydride having a particle size of about 1 μm falls between the wire meshes of the receiver plate 27 and reaches the top of the receiver plate 29. Since the UO pellets are large, the receiver remains on the tray 27.

Therefore, the UO, pellets 3 and zirconium hydride 4 shown in FIG. 6 are filled into the container 2 as shown in FIG. In this way, the spent fuel storage body 1 is obtained. The spent fuel storage body 1 is stored at a predetermined storage location.

FIG. 7 shows another example of separation of UO, pellets 3, and zirconium hydride in the hydrogenation treatment vessel 10 in FIG. 3.

Receivers made of wire mesh for sorting and separation are plates 27A and 27B.
are installed in multiple stages from above in order of coarse wire mesh. Thereby, the accuracy and efficiency of separating the UO, pellets 3, and powdered zirconium hydride 4 can be improved. Furthermore, if the receiver has a buffering function by suspending the plates 27A and 27B from the spring 30, the receiver will
It is possible to prevent the UO and pellet 3 from becoming fine powder when falling to B.

As mentioned above, according to this example, the wire mesh tray is equipped with multiple tiers of plates and has a buffering function.
This method has the advantage of improving the accuracy and efficiency of separating O, pellets and zirconium hydride.

An example of yet another sorting method is shown in FIG.

That is, by spraying gas such as hydrogen gas and/or carrier gas 31 from the nozzle 32, the wire mesh receiver removes a small amount of powdered zirconium hydride remaining on the openings of the plate 27, the UO, and the surface of the pellet 30. Blow away, catch plate 2
Move it to the bottom of 7.

This method can improve separation accuracy and efficiency.

Each of the above-mentioned sorting methods uses Zr among the constituent materials of spent fuel.
It can also be used to separate y and U Ot from other materials. That is, when considering a fuel assembly, a small amount of stainless steel bolts and nuts, or Inconel springs, etc. are included. Even if these are stored or reprocessed without separation, there is no problem since the amount is small, but it is more appropriate to separate them.

The procedure of another embodiment of the present invention is repeated based on FIG. 9, and the UOz pellet 3 and the zirconium hydride 4 are separated as described above. Thereafter, a portion of the zirconium hydride 4 is taken out from inside the hydrogenation treatment vessel 10. Hydrogenation vessel 1
The zirconium hydride 4 remaining in the tray 29 is placed on the support member 28 while remaining in the tray 29. and exhaust pump 1
3 is driven to lower the pressure in the hydrogenation treatment vessel 1O, and hydrogen is removed from the zirconium hydride 4. Moreover, the zirconium hydride 4 becomes a zirconium alloy (hereinafter simply referred to as metal zirconium).

This metal zirconium 24, UO, and pellets 3 are mixed and placed in a press container 25 shown in FIG. and,
Pressure is applied and molded using a press 26.

Since the metallic zirconium 24 is plastically deformed, a zirconium ingot in which UOt pellets 3 are mixed can be produced even when pressed with a relatively small pressure (on the order of 100 MF). Since metallic zirconium is a good conductor of heat, the decay heat generated from the UO and pellets 3 can be efficiently removed. Here, the press container 25 can also be used as the storage container 2 or the hydrogenation treatment container 10.

A molded body 22 containing a mixture of press-molded U Ot pellets 3 and metal zirconium 24 is placed in the center of a container 1 lined with a radiation shield 5 shown in FIG.
Zirconium hydride 4 is filled between the molded body 22 and the radiation shield 5.

The thus obtained spent fuel storage body 6 provides the same effects as the above-described spent fuel storage body 1. Furthermore, since metal zirconium 24 is mixed in the molded body 22, the heat conductivity is good and the decay heat dissipation effect of the UOt pellet 3 is improved.

Another embodiment of the method for manufacturing a spent fuel storage body will be described based on FIG. 12. The apparatus shown in FIG. 12 includes an exhaust pump 13A1HCt cylinder 33 and a Z
A new rO2 trap 34 is installed.

The spent fuel rods 7 are repeatedly absorbed and desorbed with hydrogen in the hydrogen treatment vessel 10 to crush the zirconium alloy member. After crushing, the zirconium alloy member becomes metallic zirconium. When this treatment is completed, the hydrogenation treatment vessel 1
0, UO, pellet 3, and golden zirconium 24 exist in a collated state. After that, it was heated to 400C by the heater 11.
The inside of the hydrogenation treatment container 10 is heated to a certain temperature, and HCt gas is introduced from the HCt cylinder 33 under that temperature condition.

7. Even when r and HO2 react, ZrC435 is generated. Since ZrC430 sublimates at about 300C or higher, it becomes a -gas and can be separated from UO and pellets 3. zr(::
430 is discharged outside the hydrogenation treatment vessel 10 by driving the exhaust pump 13 and is maintained at 3001:' or below.
C6,) Solidify and collect with wrap 34. The method of vaporizing and separating zirconium as z rct is itself known as one of the dry reprocessing methods, but in this example, the zirconium alloy member was atomized in advance to a particle size of 1 μm or less, so the reaction rate was reduced. It has the unique advantage of extremely small surface area and rapid vaporization. At this time, the chlorination reaction of the U Ot pellets can be ignored. Moreover, even if some of it becomes chloride, it will not vaporize at temperatures below 700C, so it will not mix with ZrC64.

As described above, according to this embodiment, metallic zirconium, which has been pulverized to increase the surface area, is reacted with HCl to
Rapid vaporization as c t has the advantage that metallic zirconium and UO□ pellets can be separated regardless of the presence or absence of minute fragments of UO pellets. ZrC6
Gold and A zirconium can be obtained by extracting Z r C14 from the 4-trap 34 and reducing it with Mg or Na. This metal zirconium is hydrogenated and filled into a container 2 shown in FIG. Alternatively, a part of the metal zirconium is hydrogenated and the remaining metal zirconium is mixed with U Ot pellets to form a molded product, and the molded product and zirconium hydride are filled into the container 2 as shown in Fig. 11. You may.

Another example of a separation method using vaporization is a method of separation using a heptanization reaction. That is, UO, is U F
This is a method of vaporizing and recovering it as a.

Gold MZ r y also becomes Z r F4 at the same time, but Ull
The boiling points of i', and Z r F4 are 510C and 1700C (1 atm), respectively, and below 5100, U F
Only a is vaporized. This temperature can be reduced by lowering the pressure. The separation process by fluorination reaction will be described later as an example of application to reprocessing, but in this case as well, the above-mentioned Zr
Similar to the Ct4' vaporization method, zirconium can be separated regardless of the presence or absence of UO and minute fragments of pellets.

Another method of manufacturing a spent fuel storage body will be explained with reference to FIG. 13.

In order to maintain the stability of zirconium during long-term storage, one method is to store it as zirconium hydride, rather than as metal zirconium, as described above; It is also possible to have a stable chemical form.

When producing zirconium oxide, zirconium hydride or metallic zirconium is exposed to an oxidizing atmosphere such as oxygen or air. Since the zirconium hydride or metal zirconium is fragmented or powdered as described above, the reaction proceeds quickly.

When a mixture of UO and pellets is placed in an oxidizing atmosphere, if the temperature is raised to 250C or higher, UO, becomes U, 0. and becomes powder (particle size of several 10 μm). At the time of powdering, U
O, the fission product gas trapped within the pellet is released. Also U, 0. is a stable oxidized form, and even if it is exposed to air for some reason during storage, there will be no problem of volume change due to reaction and damage to the molded product or storage container.

That is, by adding an oxidation process to Usoa, it is possible to reduce the possibility of accidents such as FP bagasse leakage or nuclear fuel material dissipation during the temporary storage period. H
,0-? Gases such as NO, etc. can promote oxidation. Below 250C, the temperature becomes U, 0A, and powdering does not occur, so removal of fission product gas cannot be expected. Once U, 0
．． It is also possible to increase the volume reduction ratio by sintering it again after powdering it. In the storage bodies shown in FIGS. 1 and 11, if excessive amounts of UO and pellets 3 are added, the amount of neutrons increases. The neutrons cannot be shielded by the zirconium hydride 4 placed in the container 2.

In such a case, the amount of neutron absorbing material may be increased.

Next, FIG. 14 shows an example when applied to a reprocessing head-end process. The spent fuel 1 is introduced into a zirconium alloy hydrogen storage step 36 and a hydrogen desorption step 37 . In this process, the fuel UOt pellets are decoated. The declad spent fuel is then sent to an oxidation step 38. In the oxidation process, in order to oxidize the highly active metallic zirconium powder, 1 to 2
The mixed gas mixed with air is kept aerated for several hours. Because of the decay heat of UO, the total temperature reaches about 400C, so the oxidation reaction progresses sufficiently. This can be confirmed by monitoring the oxygen in the argon gas at the outlet with an oxygen concentration meter at 500C. All ry changes to stable oxides. Next, the oxidation tank was heated to 5o
Raise the temperature to oc and hold for 1-2 hours. In this process, UO, U
At the same time as all of it changes to UsOa, it becomes a powder of several tens of microns. In this oxidation process 38, all gaseous fission products in the UO and pellets are removed by the fission product gas processing device 7.
sent to. Zirconium oxide and UO.

The pellets are then sent to the melting step 60 of the Beau Vtx reprocessing plant. By doing so, the dissolution conditions in the dissolution tank are significantly relaxed. That is, U3011 powder can be easily dissolved in a 6N nitric acid solution at room temperature, and the dissolution tank will not corrode. Furthermore, zirconium oxide is insoluble in nitric acid and is discharged as metal waste, and its volume is reduced in a compression and granulation step 39. The storage volume at this time is approximately 1/z6 compared to when the hull is simply stored.
becomes. On the other hand, the solution is adjusted in a filter liquid adjustment section 61 and then transferred to an extraction-separation step 62.

As described above, this example eliminates the need for a shearing process that is complicated to operate, significantly easing the dissolution conditions in the reprocessing dissolution tank, and significantly extending the life of the dissolution tank. becomes. It also has the effect of reducing the volume of metal waste.

Another embodiment applied to reprocessing is shown in FIG.

As in the previous embodiment, the spent fuel 1 is subjected to the hydrogen storage process 36.
, and after being decoated in the hydrogen desorption step 37, it is sent to a rough separation step 40, where UO, pellets, and metal Zry powder are separated. As a specific configuration of the rough separation step 40, for example, the configurations shown in FIG. 3 and FIGS. 6 to 8 can be considered. Both methods are based on separation by mesh. Detailed explanations of these are omitted since they have been described above.

The U Ot pellets separated from the metal Zry in the rough separation step 40 are transferred to the oxidation step 38 . The oxidation process in this embodiment is almost the same as in the previously described embodiment (FIG. 14), except that only air is sent to the oxidation tank.

First, the oxidation tank is heated to 8,000 ℃ while injecting air. In this process, UOt turns into powders of U, O, and F.
), the FP gas held in the UOt pellets is sent together with air to the fission product gas processing device 7 and collected. The U s Os powder is sent to a melting step 60 .

-10,000 Metal zirconium powder is discharged from the rough separation step 40, but some powdered UO is mixed in this powder. For the management of nuclear fuel materials, it is appropriate to separate and recover them.

Therefore, the powder is sent to a purification separation step 41.

A specific example of the clear separation step 41 is shown in FIG. The main equipment is a reactor 42 for fluorinating UOI, Punt and converting it into hexafluoride, a metal waste receiver 43, a fluorinated gas tank 44, a blower 45, and a reactor 42 for collecting uranium and plutonium hexafluoride. This is a cold trap 46. First, metallic zirconium powder containing a trace amount of UOt + P u 01 enters the reactor 42 through the transfer pipe 47 . The inside of the reactor 42 is filled with alumina particles to improve heat transfer, and the heat from the heater 48 on the outer wall is transferred to the powder. The reactor 42 has an alumina sintered filter 49, and the fluorinated gas is fed into the reactor from the lower part of the reactor 42 through the alumina filter 49. UO,,puo.

Since the fluorination reaction occurs at about 500C, the input to the heater 48 is controlled so that the inside of the reactor reaches that temperature. As the fluorinated gas,
Z compound (C6F), cobal 37 (CoFs), or fluorine (F2) gas can be used. 7
The fluorination reaction using nitrogen gas is as follows.

UOt +sp, =UFa +0FtPupt+5F
, =PuFa+OF'tzr-)-2F1 =7. rp
g Table 1 shows the melting points and boiling points of the main substances involved in the above reaction. From Table 1, at 500C, Z r? .

It can be seen that only Table 1 remains in the reaction vessel, and all other substances are discharged as gas. The gases discharged from reactor 42 enter cold tube 46 through conduit 50.

The cold trap 46 is maintained at approximately -60C by the cooling pipe 51, and contains uranium and plutonium hexafluoride.
Only UFs, PuFa) were collected by solidification, and Ft
, OFt is circulated, and F is the gas tank 44 and blower 4.
5 into the reactor 42 again. F filled in the system.

When the gas is consumed, it is exhausted through the conduit 52 to the fission product gas processing device 7, and the entire conduit 53 is newly passed through to fill the system with F and gas. When a predetermined amount of UP, PuFs is collected in the cold trap 46, the temperature inside the cold trap is raised to 65C using the heating tube 54. Thereby, the oxidation step 38 uses a fluidized bed. From the top of the mesh, U diluted with nitrogen gas
F.

P u F s gas is fed in, and on the other hand, heated steam containing hydrogen gas is fed in from the bottom, causing the following reaction to produce UFa and PuFe, respectively, UO! .

Convert to P u Ot.

UF',+ru+ru0=UO! +6HF'P u Fs
+)h +Ht O=P u Ot + 6HF Since all of these oxides are recovered in powder form and the quantity is small, there is no problem in sending them as they are to the melting step 4.

Further, the waste gas is hydrogen heptadide gas, which can be transferred to the FP gas processing device 7 and collected. On the other hand, zry is discharged from the reactor 42 as fluoride (ZrF4) powder. After their volume is reduced in the compression/granulation device 39 mentioned above, they are drum-canned and sent to metal waste storage.

As described above, according to this embodiment, both the previous embodiment and the
There is no need for any complicated shearing process, and the dissolution conditions in the reprocessing dissolution tank can be significantly relaxed, dramatically extending the life of the dissolution tank.

Furthermore, since metal waste does not enter the main steps of reprocessing, the process can be simplified and the volume of metal waste can also be reduced. As a reminder, compared to the previous case of discharging as oxidized Zry, when discharging as heptanized Zry, the specific gravity is about 3 (1) smaller, so the volume reduction ratio is 1 /
1.8, and the effect is somewhat weaker.

Another embodiment applied to reprocessing is shown in FIG.

In this example, after pulverizing a zirconium alloy member by hydrogen absorption and desorption, HC/ at a temperature of 400C and H from a cylinder 33 were used.
Ct gas was introduced, and the zirconium alloy member was vaporized as chloride (ZrC4) as in the previous example (Fig. 12), solidified and recovered in the cold trap 34, and subsequent steps were simplified. The place has its characteristics. That is, Zr
Since C4 sublimates at about 30 Or, it can be easily recovered in the cold trap 34. After separating and removing the zirconium alloy member, UO* is oxidized to 8o in an oxidation step 38.

After being oxidized by air under the temperature condition of C, dissolution step 6
I'm going to send it to 0.

As described above, according to this embodiment, the shearing process is eliminated;
In addition to dramatically improving the life of the dissolving tank, there is also the advantage that there is no need to newly handle metal waste.

Furthermore, another embodiment applied to reprocessing is shown in FIG. Approximately 300 tons of spent fuel lK12 in the hydrogen storage process:
' or more, hydrogen is absorbed. During this process, the zirconium alloy in the cladding material becomes brittle and fractured. In this case, the device used in the hydrogen storage step 2 is exactly the same device as shown in FIG. 3. Regarding the zirconium alloy that has been subjected to hydrogen absorption in step 2, hydrogen desorption is not performed. If you do this,
Although the zirconium alloy cannot be pulverized, the processing time can be significantly shortened compared to when an occlusion-desorption cycle is performed. The zirconium alloy that absorbs hydrogen is zrHt.
However, since this compound is quite stable at room temperature, it can be separated from UO and pellets as an insoluble residue by sending it to the dissolution step 60 in UO and pellets. Since the ZrH2 separated in the melting step 60 is flammable, it cannot be directly sent to the compression/granulation step.

In the present invention, the material is recovered as a sintered body of zirconium through the 55 pressure molding and dehydrogenation steps normally employed in powder metallurgy and the 56 sintering step. In this case, the operating condition for 55 is a molding pressure of 5 t/an''.

The dehydrogenation conditions are 110-5IllIIH at 700C.
is sufficient. The desorbed hydrogen at this time will be sent to a fission product gas processing facility for safety. Regarding No. 56, a sintered body with almost no voids was obtained by treatment at 1200C for 2 hours. On the other hand, uranium, which has become a nitrate compound during the melting process,
The plutonium solution is then sent to a filter liquid preparation step and adjusted, and then sent to the extraction-separation step 62. Also, some gaseous waste is generated in the hydrogen storage process, but most of it is generated in the melting process, so these are collectively sent to 14 fission product gas processing facilities.

As described above, by carrying out this example, only the hydrogen storage step is performed, the complicated shearing step is not required, and the conditions for dissolving the UOt pellets in the dissolving tank can be relaxed. In addition, the zirconium alloy separated in the melting process is relatively stable Z r H! Therefore, it can be recovered as a zirconium sintered body by applying conventional technology.

〔Effect of the invention〕

According to the present invention, the radiation shielding ability is high and the thickness of the radiation shielding body can be reduced.

[Brief explanation of drawings]

Fig. 1 is a vertical sectional view of a spent fuel storage body according to a preferred embodiment of the present invention, Fig. K2 is a sectional view of Fig. 1, and Fig. 3 is a sectional view of the spent fuel storage body of a preferred embodiment of the present invention. Figure 4 shows the hydrogen storage capacity of zirconium and its temperature change; Figure 8 shows the separation of UO, pellets, and zirconium hydride in the hydrogenation vessel shown in Figure 3. FIG. 9 is a flowchart showing the steps of another embodiment of the method for manufacturing a spent fuel storage body of the present invention; FIG.
The figure is used in the pressing process shown in Figure 9. Figure 12 is t4.
FIG. 3 is a system diagram of another embodiment of the apparatus shown in FIG.
Figure 8 is a process flow diagram of fuel reprocessing. 1... Spent fuel storage body, 2... Container, 3... U
O, pellet, 4...zirconium hydride, 5...
・Radiation I shield, 10...Hydrogenation treatment container, 11...
·heater.

Claims (1)

[Scope of Claims] 1. A container, a first filling region filled with spent nuclear fuel material disposed in the center of the container, and an outer periphery of the container surrounding the first filling region. a second filling region filled with zirconium hydride disposed in the spent fuel storage body. 2. Hydrogen is stored in the zirconium alloy cladding tube of the spent fuel rod, the inside of which is filled with nuclear fuel material, and then the hydrogenated zirconium alloy cladding tube is pulverized, and the nuclear fuel material and powder are pulverized. Spent fuel characterized in that the hydrogenated cladding tube that has become oxidized is separated, the nuclear fuel material is filled in the center of the container, and the hydrogenated cladding tube is filled in the peripheral area of the container. A method for producing a storage body.