JP2007033059A - Neutron shielding material and spent fuel storage cask - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a resin-made neutron shielding material having both heat resistance and cold setting property. <P>SOLUTION: The neutron shielding material contains an epoxy resin as one of its primary components 55 wt.% or more of an acid anhydride for opening and polymerizing an epoxy group and 2 wt.% or more of a curing accelerator are added to 100 wt.% of a primary agent containing a compound having at least two epoxy groups in a molecule as at least one component to carry out the reaction at a temperature not higher than 40°C to obtain a cured product. The neutron shielding material comprises the above prepared cured product. <P>COPYRIGHT: (C)2007,JPO&INPIT

Description

  The present invention relates to a neutron shielding material and a spent fuel storage container, and more particularly, a nuclear material handling facility such as a reactor vessel, a nuclear fuel reprocessing facility, a spent fuel storage facility and an accelerator facility, a radioactive material transport container, and The present invention relates to a neutron shielding material suitable for application to a radiation shielding part such as a radioactive substance storage container.

  The spent fuel assembly taken out from the nuclear reactor is cooled for a certain period in the cooling pool in the nuclear power plant to attenuate the radiation dose and the calorific value, and then transported to a treatment facility such as a fuel reprocessing plant. In recent years, overseas, fuel fuel assemblies have been transported to a centralized storage facility (dry storage facility) for storage. In order to transport spent fuel assemblies from nuclear power plants to these facilities and store them, spent fuel storage containers (metal casks) are used.

  The spent fuel storage container is provided with an inner cylinder in an outer cylinder constituting the container, and heat transfer fins made of a metal plate such as copper or aluminum having high thermal conductivity are spaced apart in the circumferential direction on the outer surface of the inner cylinder. A metal basket is provided inside the inner cylinder. Between the outer cylinder and the inner cylinder is a cured resin that is a neutron shield. The inner cylinder is a carbon steel cylinder having an upper opening and is a γ-ray shield.

  The metallic basket includes a plurality of cells, and each cell is filled with a spent fuel assembly. The metallic basket contains up to about 70 spent fuel assemblies in the case of boiling water light water reactor fuel. A primary lid for preventing leakage of radioactive material is attached to the opening of the inner cylinder, and a secondary lid is attached to the outside thereof.

  A resin, which is a neutron shield, is a substance having a high hydrogen number density containing a large number of hydrogen atoms, and a polymer compound is generally used. Of the various polymer compounds, in the spent fuel storage container, an epoxy resin is frequently used taking advantage of a good balance between heat resistance and hydrogen number density.

  In this case, a liquid epoxy base material, a curing agent, aluminum hydroxide that imparts flame retardancy, and boron carbide that is a neutron absorber are mixed uniformly, and the above inner and outer cylinders are mixed. ， Inject into the space surrounded by heat transfer fins and cure at room temperature.

  When using a room temperature curing type epoxy resin as a neutron shielding material, it is usual to use an amine compound, particularly an aliphatic or alicyclic amine, as a curing agent with respect to the epoxy main agent (for example, patent documents) 1).

  On the other hand, a thermosetting epoxy resin may be used for the neutron shielding material (see, for example, Patent Document 2). In this case, an amine-based curing agent or an acid anhydride and a curing accelerator are added to the epoxy main agent, and a cured product cured at a temperature higher than normal temperature is applied to the neutron shielding material. When heat-cured, the crosslink density of the resin increases, and excellent heat resistance is obtained.

Japanese Patent Laid-Open No. 2001-108787 (pages 4-5, FIGS. 1-3) Japanese Patent Laying-Open No. 2003-167091 (FIG. 1 on page 6)

  In order to provide sufficient storage for spent fuel assemblies in the cooling pool, dry storage inside or outside the nuclear power plant is being studied. In the future, the possibility of dry storage of spent fuel assemblies with a short cooling period in the cooling pool, and the higher fuel burnup (> 45 GWd / ton) spent fuel assemblies will be dry-stored. There is also a possibility.

  The spent fuel assembly having a short cooling period in the cooling pool and the spent fuel assembly of the high burnup fuel assembly generate a large amount of heat when the fission product nuclide and the transuranium element decay. When such a spent fuel assembly is stored in a spent fuel storage container, if the number of bodies stored per spent fuel storage container is increased, the neutron shielding material having a lower thermal conductivity than metal is applied. Heat load increases.

  An object of the present invention is to provide a neutron shielding material that has both heat resistance that can withstand use under higher temperature conditions and room temperature curability that can apply the direct filling method that has been widely used for resin construction, and its neutrons. It is an object of the present invention to provide a spent fuel storage container that uses a shielding material to increase the storage efficiency of highly exothermic fuel.

  In order to solve the above problems, the present invention provides a neutron shielding material having an epoxy resin as one of the main components, the main component containing at least one compound containing two or more epoxy groups in the molecule as the epoxy group. A neutron shielding material composed of a cured product obtained by adding an acid anhydride for ring-opening polymerization and a curing accelerator and reacting at a temperature of 40 ° C. or lower is proposed.

  The neutron shielding material of the present invention is made of a general-purpose material, has heat resistance that can withstand use at higher temperatures, and room temperature curability for applying a direct filling method that has been widely used for resin construction. Have both. When this neutron shielding material is applied to a spent fuel storage container, the number of highly exothermic fuel storage bodies can be increased.

  Next, with reference to FIGS. 1-3, the Example of the neutron shielding material and spent fuel storage container by this invention is described.

  In the description of the present embodiment, normal temperature refers to a temperature of 40 ° C. or lower.

  An acid anhydride for ring-opening polymerization of an epoxy group and a curing accelerator for accelerating the reaction between the epoxy group and the acid anhydride were added to an epoxy main agent containing at least one compound containing two or more epoxy groups in the molecule. Later, a neutron shielding material was composed of a cured product obtained by allowing to stand at room temperature.

  Even when the neutron shielding material of the present invention is heated at 150 ° C. to 170 ° C., the degree of decrease in the hydrogen number density is extremely small, so that the neutron shielding performance does not deteriorate during service. The spent fuel storage container to which the neutron shielding material of the present invention is applied can increase the number of spent fuel assemblies loaded with a spent fuel assembly or a high burnup fuel assembly with a short cooling period in the cooling pool. it can.

  The inventors clarify the problem when the spent fuel assembly having a short cooling period in the cooling pool and the spent fuel assembly of the high burnup fuel assembly are loaded into the spent fuel storage container. At the same time, various measures to solve the problem were examined. The examination process will be described.

  Spent fuel assemblies with a short cooling period in the cooling pool, and spent fuel assemblies of the high burn-up fuel assembly, have a large calorific value due to the decay of fission-generated nuclides and transuranium elements. It has been found that the temperature of the neutron shielding material may reach a maximum of about 170 ° C. when the number of bodies stored per spent fuel storage container of the fuel assembly is increased.

  When the temperature rises, the neutron shielding material mainly composed of a polymer compound gradually decomposes due to thermal oxidation degradation due to heat and oxygen or radiation degradation due to γ rays or neutrons, and consumes hydrogen atoms. As the hydrogen atoms are consumed, the neutron shielding performance gradually decreases. The higher the temperature, the greater the rate at which hydrogen atoms are lost due to deterioration.

  A spent fuel assembly with a short cooling period in the cooling pool and a spent fuel assembly of the high burnup fuel assembly (referred to as a highly exothermic spent fuel assembly) are stored at high density in the spent fuel storage container. Therefore, it is desired to develop a neutron shielding material in which the loss of hydrogen atoms is slow and the shielding performance does not deteriorate even when used at the high temperature for a predetermined period. If the loss rate of hydrogen atoms at the above temperature of the neutron shielding material is lower than the decay rate of the neutron emission nuclide in the spent fuel assembly, the radiation dose on the spent fuel storage container surface can be suppressed low.

  From these viewpoints, one of the issues for high-density storage of high-heat-generated spent fuel assemblies is a polymer compound with a high hydrogen number density, and loss of hydrogen atoms occurs when used under high temperature conditions. It is to produce a neutron shielding material using a difficult polymer compound.

  The inventors have made various studies toward the realization of an epoxy-based neutron shielding material in which the curing reaction proceeds at a sufficient rate even at room temperature and the loss of hydrogen atoms hardly occurs even when the cured product is heated to about 170 ° C. .

  Generally, when curing epoxy at room temperature, an amine curing agent is used. On the other hand, for applications other than shielding materials, it is common to use acid anhydrides as curing agents.

  Thus, the inventors have conducted experiments and research under the policy of curing the epoxy at room temperature with a more general-purpose acid anhydride, and as a result, have found a compounding ratio of an epoxy resin having both room temperature curing and heat resistance. The process will be described.

  FIG. 1 is a diagram showing the dynamic viscoelasticity measurement results of a cured resin product prepared by changing the addition ratio of the curing agent to the alicyclic epoxy main agent. FIG. 2 is a diagram showing dynamic viscoelasticity measurement results of a cured resin product prepared by changing the addition ratio of the curing accelerator while keeping the addition ratio of the curing agent constant with respect to the same alicyclic epoxy. .

  1 and 2 are theoretically functions of the average molecular weight between the crosslinking points of the epoxy resin, and the larger the value, the denser the crosslinking of the resin.

  As a result of the test, when 55 to 85% by weight of the curing agent and 2 to 5% by weight of the curing accelerator are added to 100% by weight of the epoxy main agent, the epoxy is cured at a temperature of 40 ° C. or less. From this point of view, it was found that the crosslinked state was sufficient. The cured product sample thus obtained was heated for 1000 hours or more. As a result of analyzing and evaluating the sample after heating, it was predicted that the decrease in shielding performance could be kept small.

  The reason why the room temperature is defined as 40 ° C. or less is that when the temperature exceeds 40 ° C., the reaction proceeds so much that it may be difficult to manufacture.

  When an epoxy resin is used as a neutron shielding material, a metal oxide hydrate or the like may be added as a flame retardant in order to impart flame retardancy. In addition, a neutron shielding material mainly composed of a polymer material may add a boron compound having a large neutron absorption cross-section as a neutron absorbing material in order to supplement neutron shielding performance.

  For example, a shielding material used in a spent fuel storage container is a typical example. Usually, aluminum hydroxide or magnesium hydroxide is added as a flame retardant, and boron carbide is added as a neutron absorber.

  Inventors examined various mixing ratios of a flame retardant and a neutron absorber. As a result, when the flame retardant is 50 to 65% by weight and the neutron absorber is 1 to 10% by weight, the objectives of imparting flame retardancy and enhancing neutron shielding performance, which are the original objectives, are sufficiently achieved, and 40 ° C. It has been found that excellent filling workability can be realized at the following temperatures.

  That is, it was confirmed that the viscosity of the liquid resin mixture continued not to exceed 150 dPa · s for 2 hours or more after blending as well as immediately after blending.

  According to the present invention, a resin obtained by mixing an epoxy main agent, an acid anhydride, a curing accelerator, a flame retardant, and a neutron absorber and curing at a temperature of 40 ° C. or less is applied to the neutron shielding material by the inventors' own experiments and research. It was obtained as a result of clarifying what can be done.

Resin, which is a neutron shield, is a substance with a high hydrogen number density that contains a large number of hydrogen atoms. From the viewpoint of neutron shielding ability, the blending ratio of each component is preferably such that the hydrogen number density in the cured product of the main agent and the additive is 5 × 10 ²² hydrogen atoms / cm ³ or more.

  If necessary, a metal hydride or a hydrogen storage alloy may be added to the main agent.

  In this example, a neutron shielding material using a cured product obtained by curing an alicyclic epoxy main agent at a temperature of 40 ° C. or lower using an acid anhydride and a curing accelerator will be described. The neutron shielding material of the present example is manufactured as follows.

  As the main agent, an alicyclic epoxy main agent having an epoxy equivalent of about 240 g / eq is used. About 100% by weight of a maleic anhydride adduct of methylcyclopentadiene as a curing agent and 2% by weight of 2-ethyl 4-methylimidazole as a curing accelerator are added to 100% by weight of the main agent.

  Further, as a flame retardant, 170% by weight of aluminum hydroxide having an average particle diameter of about 10 μm is added to 100% by weight of the main agent. Further, about 4% by weight of boron carbide powder having an average particle size of 45 μm or less is added to 100% by weight of the main agent.

  The above components are sufficiently kneaded and homogenized for about 30 minutes, and then defoamed by reducing the pressure to about 10 torr.

  The structure of the spent fuel storage container filled with the resin after degassing will be described.

  FIG. 3 is a perspective view showing an example of the structure of a spent fuel storage container to which the neutron shielding material of the present invention should be applied.

  The spent fuel storage container 1 includes an outer cylinder 3 and an inner cylinder 2, and the outer cylinder 3 and the inner cylinder 2 are coupled by heat transfer fins 4.

  The heat transfer fins 4 are attached at intervals in the circumferential direction, and transmit heat inside the spent fuel storage container 1 from the inner cylinder 2 to the outer cylinder 3 and dissipate heat from the surface of the outer cylinder 3 to the atmosphere. .

  The inner cylinder 2 is a carbon steel structure having an upper opening, and also has a gamma ray shielding function. Inside the inner cylinder 2, a metal basket 7 formed in a lattice shape is installed. The metal basket 7 includes a plurality of cells, and a spent fuel assembly is accommodated in each cell.

  A primary lid 8 for preventing leakage of radioactive material is attached to the opening of the inner cylinder 2, and a secondary lid 9 is attached to the outside thereof.

  The spent fuel storage container 1 is handled using an upper trunnion 10 and a lower trunnion 11 provided on the side surface of the outer cylinder 3.

  A neutron shield 5 is disposed in each space formed by the heat transfer fin 4 between the outer cylinder 3 and the inner cylinder 2. The primary lid 8 attached to the opening of the inner cylinder 2 is also filled with neutron shields 6 respectively.

  The defoamed liquid resin is pumped from the kneading vessel and filled into the spaces formed by the heat transfer fins 4 between the outer cylinder 3 and the inner cylinder 2. The resin is cured in about 2 days at a temperature of 40 ° C. or lower.

  In addition to the construction method in which the neutron shielding material of the present invention is directly filled between the inner and outer cylinders as described above, a construction method in which a block-shaped molded body is once produced and inserted between the inner and outer cylinders can be employed. In the direct filling method, the amount of shielding is confirmed by inspecting the cross-sectional area of the injection site and injecting height, whereas in the block mounting method, the amount of shielding is confirmed by inspecting the size and weight of the block.

  According to this example, a resin for neutron shielding material having both heat resistance that can withstand use at a higher temperature and room temperature curing that can apply the direct filling method that has been widely used in resin construction has been obtained. . When this neutron shielding material is applied to a spent fuel storage container, the number of highly exothermic fuel storage bodies can be increased.

It is a figure which shows the dynamic viscoelasticity measurement result of the resin cured material prepared by changing the addition ratio of the hardening | curing agent with respect to an alicyclic epoxy main ingredient. It is a figure which shows the dynamic viscoelasticity measurement result of the resin cured material prepared by changing the addition ratio of a hardening accelerator variously, making the addition ratio of a hardening | curing agent constant with respect to the same alicyclic epoxy. It is a perspective view which shows an example of the structure of the spent fuel storage container which should apply the neutron shielding material of this invention.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 Used fuel storage container 2 Inner cylinder 3 Outer cylinder 4 Heat transfer fin 5 Neutron shield 6 Neutron shield 7 Metal basket 8 Primary lid 9 Secondary lid 10 Upper trunnion 11 Lower trunnion

Claims (12)

In neutron shielding material with epoxy resin as one of the main components,
With respect to 100% by weight of a main agent containing a compound containing two or more epoxy groups in the molecule as at least one component, 55% by weight or more of an acid anhydride curing agent for ring-opening polymerization of the epoxy group and 2% of a curing accelerator A neutron shielding material comprising a cured product obtained by adding at least 40% and reacting at a temperature of 40 ° C. or less. In the neutron shielding material according to claim 1,
The main material is a material obtained by mixing or mixing a bisphenol A type epoxy resin, a novolac type epoxy resin, an alicyclic glycidyl ether type epoxy resin, various glycidyl ester type epoxy resins, and a glycidyl amine type epoxy resin. And
The acid anhydride curing agent is maleic anhydride, phthalic anhydride, methyl nadic anhydride, succinic anhydride, trimellitic anhydride, chlorendic anhydride, or a material obtained by mixing these compounds alone or in combination.
The neutron shielding material, wherein the curing accelerator is an imidazole compound. In the neutron shielding material according to claim 1 or 2,
A neutron shielding material comprising a flame retardant added to the main agent. In the neutron shielding material according to claim 3,
The flame retardant is a metal hydroxide such as magnesium hydroxide, aluminum hydroxide or calcium hydroxide, a hydrate of the metal oxide, an inorganic phosphate compound such as ammonium polyphosphate, or a halogen compound such as tetrabromobisphenol A. The neutron shielding material characterized by including either of these. In the neutron shielding material according to claim 4,
When a metal hydroxide or a hydrate of the metal oxide is used as the flame retardant, the flame retardant is 50 to 65% by weight based on the main agent. In the neutron shielding material according to claim 4 or 5,
A neutron shielding material, wherein the flame retardant has a blending ratio such that an oxygen index of the cured product exceeds 21. In the neutron shielding material according to any one of claims 1 to 6,
A neutron shielding material, wherein a neutron absorber is added to the main agent. In the neutron shielding material according to claim 7,
The neutron shielding material, wherein the neutron absorbing material contains any one of a boron compound, a cadmium compound, a gadolinium compound, and a samarium compound. The neutron shielding material according to claim 8,
When boron carbide or boron nitride is used as the neutron absorber, the neutron absorber is 0.1 to 10% by weight based on the main agent. In the neutron shielding material according to any one of claims 1 to 9,
A neutron shielding material, wherein a metal hydride or a hydrogen storage alloy is added to the main agent. In the neutron shielding material according to any one of claims 1 to 10,
The mixing ratio of each component is a mixing ratio such that the hydrogen number density in the cured product of the main agent and the additive is 5 × 10 ²² hydrogen atoms / cm ³ or more. The outer cylinder and the inner cylinder are connected to each other by heat transfer fins attached at intervals in the circumferential direction, and a metal basket for storing the spent fuel assembly in each cell is installed inside the inner cylinder. In the spent fuel storage container with a lid that prevents leakage of radioactive material at the opening of the cylinder,
The spent fuel storage container characterized by filling each space formed with the said outer cylinder, the said inner cylinder, and the said heat-transfer fin with the neutron shield as described in any one of Claim 1 thru | or 11.

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