JPS5941156B2 - Spent fuel transfer pot - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a cooling pot (transfer pot) used when removing spent nuclear fuel from a reactor.

When taking spent fuel out of the reactor, the fuel assembly is first stored in a transfer pot using the in-core relay mechanism, then transferred as is to the out-of-core relay container using an in-core cask car, and then stored using the out-of-core relay mechanism. Store it in a rack.

During this process, the fuel assembly continues to have decay heat removed by the sodium in the transfer pot.

In a conventional transfer pot, for example, as shown in FIG. 1, there are cooling fins 2 on the outer periphery of the pot body 1 of the device, the inside is filled with sodium 3, and the spent nuclear fuel assembly 4 is placed in a trumpet tube. It is inserted so as to be immersed in the sodium 3 while still in the entrance nozzle 5, and to secure a gap at the bottom of the entrance nozzle 6 to allow the sodium 3 to easily flow.
The structure is such that the trumpet tube 5 is supported by a support mechanism 7.

The fuel assembly is divided into three regions in the axial direction from the lower entrance nozzle 6 side: a lower blanket, a core heating section, and a blanket.

The sodium 3 is heated by the decay heat of the core heat generating part, circulates in the direction of the arrow by natural convection, and radiates heat through the cooling fins 2.

Since no heat is supplied to the sodium at the entrance nozzle, there is a possibility that the sodium that is cooled by the cooling fins 2 and descends will solidify there (below 98° C.).

If it solidifies, the sodium 3 will no longer circulate inside and outside the wrapper tube 5, and the decay heat removal effect of the core heat generating section will be impaired.

Also, if it solidifies, when removing the fuel assembly,
It becomes necessary to dissolve the sodium, which requires a heating mechanism, which complicates the structure of the pot, increases costs, and
Unfavorable aspects such as a decrease in reliability arise.

In order to prevent sodium from solidifying in the lower part of the transfer pot, a method has been proposed in which an insulating space is provided around the lower part of the pot, but this is not preferable from the viewpoint of actively removing heat.

Additionally, there is a good chance that some of the fuel pins will have broken cladding.

This has been experienced in advance reactors of sodium-cooled fast reactors overseas.

When the cladding ruptures, some of the fission products and fuel inside are released into the sodium.

Among the fission products, cesium (134C, 137
Cs) has a long half-life, is easily soluble in sodium, and at the same time is more volatile than sodium, making it easy to become a source of radioactive contamination.

Therefore, in the process of handling the spent fuel mentioned above and in the process of subsequent fuel processing, 134C,,137Cs
It is necessary to prevent the spread of contamination.

However, the above-mentioned conventional transfer pots do not have sufficient countermeasures for this problem.

In other words, there is no function to actively suppress the presence of 134C5, 137C8 in the molten sodium and the sodium cover gas. death,
The natural circulation of sodium in the pot is maintained without sacrificing heat removal efficiency, increasing the reliability of decay heat removal, simplifying fuel handling and saving labor, and reducing the expected amount of fission products. An object of the present invention is to provide a spent fuel transfer pot that can suppress the spread of radioactive contamination, improve safety and maintainability, and improve work efficiency.

In order to achieve this object, the present invention uses pyrolytic graphite, which is a kind of carbon material, around the fuel installation part in the pot.
(hereinafter abbreviated as IT PG II).

Hereinafter, the present invention will be explained in detail based on the drawings.

FIG. 2 is a longitudinal cross-sectional view showing one embodiment of the present invention, and FIG. 3 is a cross-sectional view taken along line II.

There are cooling fins 2 on the outer periphery of the pot body 1, and the inside is filled with sodium 3.

The spent nuclear fuel assembly 4 is inserted into the trumpet tube 5 so as to be immersed in the sodium 3.

The trumpet tube 5 is supported by a support mechanism 7 so as to secure a gap below the entrance nozzle 6 to allow the sodium 3 to flow easily.

The configuration up to this point is the same as the conventional transfer pot.

A feature of the present invention is that PG (pyrolytic graphite) 10 is disposed around the trumpet tube 5.

PG is a type of carbon material with highly anisotropic physical properties, and when it is heat-treated at about 2900℃, its thermal conductivity in one direction (a-axis direction) is higher than that of copper. It is known to have better insulation properties than insulating bricks in the other direction (C-axis direction), and is also called pyrolytic graphite, vapor phase graphite, or oriented graphite depending on the manufacturing method.

It is suitable for transporting heat between certain parts.

In the present invention, PGlo is attached along the central axis so that its longitudinal direction is in the a-axis direction, that is, in a direction with good thermal conductivity.

As shown in FIG. 3, in this embodiment, a large number of PGIOs are rod-shaped and arranged in parallel by bands 11ε and support plates 12.

At this time, the upper end surface of PGlo is located near the lower end of the fuel heat generating section where the temperature of the sodium 3 is high, and the lower end surface is located near the top of the entrance nozzle 6 of the fuel assembly.

Note that the shape of PGlQ is not limited to a cylindrical shape,
The shape may be a plate or a hollow cylinder, but the shape is arranged so as not to obstruct the downward flow of the sodium 3.

PG has corrosion resistance against molten metal and mechanical strength equivalent to ceramic, so there is no need to attach a special protective cover when installing it in sodium.
Can be easily and compactly stored in a transport pot.

There are no maintenance difficulties. Even if the transfer pot of the present invention is
The basic function of removing decay heat through natural convection of internal sodium is the same as in conventional systems.

Furthermore, in the present invention, PGlo also adds the effect of effectively transporting heat from the high temperature section to the low temperature section where sodium solidification is a concern, thereby preventing sodium assimilation in the low temperature section, which is a drawback of the conventional technology. Natural circulation of sodium can be maintained without reducing heat removal efficiency.

In addition, if it is desired to widely warm the low temperature region in the vertical direction, this can be easily achieved by dispersing the positions of the lower end surfaces of the PGlo in the vertical direction.

In addition, the spread of contamination of fission products cesium (134C, 137C,), which is another drawback of the conventional technology, is also caused by this PGl
It is possible to deter them by capturing them at o.

In other words, graphite has 137C in sodium.
, has been confirmed through experiments by many researchers.

Since PG used in the present invention is also a type of graphite, it naturally has the property of collecting this 134.137C8.

It is also known from experiments that even if 137Cs is adsorbed in one phase on stainless steel in sodium, almost all of it will be transferred to this if graphite is added.

For this reason, according to the present invention, the fuel released due to the breakage of the fuel pin cladding tube and adsorbed to the fuel assembly 4 and the trumpet tube 5
34, 137 C8 is transferred to PGlo while cooling in the transfer pot, which greatly facilitates subsequent handling of the fuel assembly 4 and wrapper tube 5.

Therefore, good secondary effects also occur in terms of safety, efficiency, and economy.

Since the present invention is a spent fuel transfer pot constructed in this manner, it reliably maintains the natural circulation of sodium within the pot without impairing heat removal efficiency, increases the reliability of decay heat removal, and transfers fuel to the spent fuel. In addition to simplifying and labor-saving the handling of nuclear fission products, it also prevents the expected spread of radioactive contamination from fission products, improving safety and maintainability, and increasing work efficiency. It has excellent effects.

[Brief explanation of the drawing]

FIG. 1 is an explanatory diagram of the prior art, FIG. 2 is a vertical sectional view of an embodiment of the transfer pot according to the present invention, and FIG.
It is an I sectional view. 1... Pot body, 2... Cooling fin, 3...
Sodium, 4... Fuel assembly, 5... Trumpet tube, 10... PG, 11... Band, 12... Support plate.

Claims (1)

[Claims] 1 A device for transferring spent nuclear reactor fuel from the inside of the reactor to the outside of the reactor. The high pyrolytic graphite is installed along the central axis, and its mounting position and dimensions are such that the top end surface of the pyrolytic graphite is located in the high heat generation area and the bottom end surface is located in the non-heat generation area. A spent fuel transfer pot featuring: