Fuel rod and method of making pellets for said fuel rod

FIELD: physics, atomic power.

SUBSTANCE: invention relates to nuclear reactor fuel rods. Fuel rod cladding has an elliptical cross-section. Each nuclear fuel pellet along the longer axis of the cladding has a truncated elliptical shape, and the shorter axis of the pellet has the same length as the shorter axis of the cladding, minus the fitting gap j, wherein the difference in length of the longer axis of the cladding and the truncated longer axis of the pellet is much greater than said fitting gap j. The invention also relates to a method of making nuclear fuel pellets and a method of packing said pellets, which enables to form a fuel rod.

EFFECT: lower probability of deformation of fuel rod cladding and release of fission products into the coolant.

9 cl, 3 dwg

 

The technical field to which the invention relates

The invention relates to a new type of fuel rod.

Prospective application of a new type of fuel rod include nuclear water-water energy reactor (WWER) and gas-cooled nuclear reactors on fast neutrons (GBAR), called reactors of the 4th generation.

For all purposes of this application the term "nuclear reactor" refers to the generally accepted meaning of the term, currently used, namely power plants, producing energy based on fission of nuclei with the help of fuel elements, in which the fission cores that emit energy in the form of heat the energy extracted from the fuel cell by heat exchange with the coolant, which cools these fuel elements.

For all purposes of this application, the term "fuel rod" refers to the official meaning of the term, the definition of which is given, for example, in the dictionary of nuclear physics and technology, namely, the small diameter tube (or narrowed), plugged at both ends, forming a fuel element of a nuclear reactor containing the nuclear fuel. With this arrangement, formed fuel element of nuclear fuel for which in the present description every�retenu used preferred term is "fuel rod".

The invention thus discloses a new design of fuel rods, having improved thermomechanical properties during mechanical interactions between the nuclear fuel tablets and sheath.

The level of technology

There are various types of fuel fuel elements used depending on the operating parameters and characteristics of the reactor. The so-called power plants of the 3rd generation and, in particular, water-water energy reactor (WWER) use fuel rod type elements with round cross-section.

The inventor has set a goal to improve the design of fuel elements and started with a desire to understand the design principles of all known fuel elements of the reactor and to identify their functional limitations.

Fuel element shall exercise the following basic qualities:

the density of fissile atoms must match the parameters of the effect of neutrons and the energy density per unit volume reaction volume

it needs to capture the warmth between the nuclear fuel and coolant

it should hold solid and gaseous nuclear fission products emitted by the fuel in the reactor. Reaction of fission, p�oshidashi inside of nuclear fuel, generate solid and gaseous fission products, which determine the likely significant swelling of nuclear fuel. The process of swelling, particularly of the gas swelling, is activated by heat, which activates the mechanisms through which gaseous fission products are at the limits of nuclear fuel. It is therefore necessary that the fuel cladding was able without loss of integrity to compensate for these deformations and gaseous emissions from fuels.

The intensity of nuclear fission reactions inside the fuel directly correlated with the magnitude of the heat capacity per unit volume, which should be given to the coolant through the fuel cladding.

It is therefore necessary to minimize thermal resistance between the heat source and a cooling fluid in order to limit the maximum fuel temperature and the effects caused by thermal flow: gradient in nuclear fuel and various extensions of the fuel and shell.

The density of fissile material (nuclear fuel) in a reaction volume depends mainly on the shape of fuel elements, which limits their ability to be placed in a given volume with the desire to maximize the fill factor of e�CSO volume while maintaining at the same time, the required bandwidth for the refrigerant, to ensure the removal of heat released heat-generating elements, at an acceptable pressure loss.

The main heat-generating elements traditionally used in nuclear power plants, can be classified into three types, namely the element plate type (all forms), the cylindrical element type, elongated in the direction of the axis (usually having a round or circular cross-section), which forms an element of the rod, and a spherical type element, usually in the form of particles of small diameter (approximately equal to one millimeter).

Furthermore, the known composite fuel elements formed from spherical particles encapsulated in inert matrix existing in three shapes, namely, bowls, plates and compact form that is used in high temperature reactors (TUE).

Each of these three types of fuel elements combines various solutions to emerging problems, and the necessary compromise between the existing variants of design solutions for the application. The scope of each fuel element is in practice limited by the characteristics of the chosen design.

So, records are made with a shell that behaves like a shell with a high degree of splyusnutoi (the ratio�the communication between the free length of the shell and its thickness).

Due to its plasticity the geometric shape of the shell material adapts itself to the geometry of the Central part of the nuclear fuel. This means that different transverse deformation in nuclear fuel and cladding swelling and stretching) can be compensated with very low levels of mechanical stress. However, the lamellar structure has insufficient capacity to contain the strain imposed on her by the nuclear fuel in the thickness direction, due to the very low stiffness of the shell in the transverse direction to its surface. This freedom allows fuel to deform anisotropic and preferably in the same direction. In addition, the design is highly unstable under the action of bending stresses in the case when in the plane of the placement of the design, the entire design, or locally (e.g., hot spot) applied compression force, especially in cases in which the fuel core fuel element is not attached to the shell (not in contact) or there is only a mild contact.

Good thermal contact between the nuclear fuel and the cladding is required to maintain fuel temperature within an interval sufficiently low temperatures so that the fuel is not allocated gaseous fission products �ri all modes of reactor operation.

Therefore, plate fuel elements used only for cold nuclear fuels, in other words, in the temperature range in which the nuclear fuel does not emit gases, and at moderate levels of thermal energy per unit volume.

Parameters optimization of plate elements for a given power per unit volume usually includes the thickness of the plate and the quality of the contact between fuel and cladding corrosion control membrane and absence of deterioration in its ability to plastic deformation in the process.

The main types of violations of the plate elements are, or loss of plasticity of the shell under the action of applied deformation (corrosion deterioration or radiation hardening), or increase thermal resistance between the nuclear fuel and the coolant (for example, due to the formation of corroded areas on the shell that creates thermal resistance, exfoliation fuel from the shell with the formation of a gap due to local buckling of the shell), causing a temperature rise of nuclear fuel, the release of gaseous fission products and the increase of the internal pressure inside the shell, leading to destruction of the shell due to uncontrolled deformation.

Cylindrical fuel elements are� a, for example, cylindrical containers with nuclear fuel used in grafitowych reactors, the rods used in water-water power reactors (VVER), or fuel cells rod-type nuclear reactors on fast neutrons (BR).

Construction of these cylindrical elements characteristic of the presence of a radial gap between the nuclear fuel in the form of tablets and sheath, inside which laid these pills with the formation of the post that allows you to compensate for different deformation between the nuclear fuel and the shell; this gap is capable of, at least, to compensate for the different expansion during the first lift-power heat-generating element and part of the swelling of the fuel, which cannot be rezorbirovanny themselves at the expense of fluidity and re-seal in its internal cavities, in other words, in the cavities formed by the Central hole and the pores in the fuel. Nuclear fuel must also operate at the temperature at which it can activate the compensatory mechanisms of deformation.

On the other hand, it allocates a certain amount of gaseous fission products of the fuel.

The second expansion volume formed inside the shell at the end of the column of nuclear fuel tablets (fuel pellets) in order to limit the amount of internal d�effect in the fuel element.

The main optimization parameters of these cylindrical elements are the initial radial clearance between the nuclear fuel and the cladding, in other words, the radial Assembly gap, the quality of thermal coupling between the nuclear fuel and the shell with a fluid medium (gas seal or a seal of molten metal), the actual density of the cross section of the shell nuclear fuel, as defined radial clearance, pores, voids, including a Central hole and/or lenticular cavities in the ends of the tablets, the rigidity of the shell (thickness), mechanical properties (ultimate strength and the ability to plastic deformation) and patterns of behavior of the shell, and nuclear fuels (buckling and creep).

The radial clearance between the tablet and the shell is filled with gas, and the thickness of the shell to form a thermal resistance in the radial direction, which determines the transfer of heat between the coolant and the nuclear fuel tablets.

thermal resistance in the process of change as there is a change in a radial gap and the deterioration of thermal conductivity due to emission of gaseous fission products. The change in thermal resistance complicates the control of the maximum temperature of the fuel, which requires that�tsya fact, nuclear fuel should not reach its melting temperature under all conditions of operation. In addition, the use of an element of this type in the "chamber pressure" involves the use of a material capable of retaining the fuel element in position without the risk of a sudden rupture (instantaneous and/or delayed) under the action of pressure. To achieve this result is usually used round section, because it better resists the action of pressure. Thus, in a situation of mechanical interaction between fuel and sheath, the sheath being under the influence of ring strain, exhibits a high ring stiffness. As a result, the fuel in two radial directions is kept from moving, and only in the axial direction partially has freedom, and that partial freedom is dependent on adhesion between the pellets and the sheath.

Circumferential pressure that acts on the fuel side of the membrane, activates acting on the fuel redistribution mechanisms, in other words, overcrowding.

Consequently, the choice of the shell material is of particular importance because it must provide sufficient ultimate tensile strength in a predetermined range of operating temperatures, the ability to plastic deformation�AI in plastic and thermal deformation and sufficient ultimate strength, usually a component of more than 20 MPa.√m in the temperature range corresponding to the entire area, which employs fuel cells. Therefore, the limiting operating parameters of these elements (temperature and power per unit volume) is set by selecting the shell (short tear resistance and creep strength depending on temperature) and nuclear fuel (melting point).

The main form of residual fracture associated with the type of fuel element is short-term mechanical interaction between fuel and sheath, exceeding the ability of the shell to deform, for example, in situations in which the reactor power increases to a higher level compared with the previous work level, or in such working conditions in which the fuel temperature will not activate its own mechanisms of self-cancellation of deformations or enables them only slightly.

Finally, in spherical elements, such as representing particles used in high temperature reactors (TUE), covering various layers are sequentially deposited on the fissile core, which should be centered. This is achieved through the creation of voids in the form of pores within the fissile core and prom�safely layer, called "buffer" that has a very high porosity, which preserves the continuity between the fissile core and shell layers.

Different deformation between the nuclear fuel and the cladding, in other words, the covering layer are compensated by the voids; during operation, the gradual compaction of the buffer under the action of neutron flux frees the radial clearance, which prevents a strong mechanical interaction between fissile core and covering layers. In addition, the internal volume within the shell retain gaseous fission products emitted fissile nuclear fuel. The spherical shape of the shell, therefore, is well adapted to withstand the increasing internal pressure.

The optimization parameters of elementary particles are essential when choosing the material (type, structure, properties and regularities in the behavior under the action of neutron flux and temperature) and the thickness of the various layers.

It should be noted that the spherical fuel elements are used only in high-temperature and gas-cooled reactors (BP).

Their main mode of residual fracture corresponds to the strong interaction between fissile core and covering layers (creating mechanical stresses in a given deformational), which can cause the destruction of the protective sheath; based on this spherical shell is the worst form of shell because it doesn't leave directions for the deformation of nuclear fuel (in addition to maximum seal) to weaken the strength of interaction (creation of hydrostatic pressure in the internal volume of the shell).

Considered the type of the spherical fuel element is also used in various composite forms in which particles are dispersed in the matrix through which the heat transferred to the coolant, with a very low content of nuclear fuel in the reaction volume of the reactor, of the order of several percent per unit of volume. In addition, this design reduces the risk of failure envelope at high intensity of combustion (or burning) of nuclear fuel.

Subject to the foregoing, the inventor believes that each of the three types of fuel elements has advantages that can be summarized as follows:

plate characterized by good heat transfer and good adaptability during the mechanical interaction between the nuclear fuel tablets and sheath,

cylindrical fuel elements (rods) and spherical elements have good resistance to the pressure of the gaseous prod�Chow division.

On the other hand, given the above, one can also conclude that the major shortcoming of currently used elements of cylindrical type (rods) is that their thermo-mechanical characteristics in the implementation of the mechanical interaction between the nuclear fuel tablets and the shell can be controlled.

In this regard, the author of the invention as the main objectives set for the improvement of thermomechanical characteristics of the fuel element rod type, working in conditions of interaction between the nuclear fuel tablets and shell, currently used in reactors of the 2nd and 3rd generations.

Proposed new fuel elements can also be used for nuclear fast reactors gas-cooled 4-th generation.

A more General object of the invention is to create a fuel element rod type, which combines the advantages of various types of existing fuel elements of the kind referred to above, and in order to become possible to satisfy the following technical requirements:

1) reaching the values of relative mass of fuel per unit of volume equal to the values implemented in the existing rods with a circular cross section,

2) ensure�ood optimal transfer of heat from the nuclear fuel tablets to the coolant during the entire lifespan, thus, progress indicators should be comparable to the transmission of heat when using plate elements (heat exchange, preferably on two opposite sides),

3) the elimination of the danger of destruction of the shell by controlling the mechanical interaction between the nuclear fuel tablets and sheath.

Another object of the invention is to provide a fuel element rod type, method of construction which does not meet fully the industrial equipment installed for the manufacture of existing fuel element rod type, having a circular cross-section.

Disclosure of the invention

The task is solved in the fuel rod, located in the longitudinal direction containing large amounts of nuclear fuel tablets, stacked on top of each other, and the shell is made of a material transparent to neutrons surrounding the column of tablets, in the fuel rod in cross-section, transverse to the longitudinal direction,

- the shell has an elliptical shape, and the inner surface has a greater axis with length 2×a and the minor axis with length 2×b,

- each tablet of nuclear fuel has generally the shape of an ellipse, truncated at the ends of the major axis of the shell, the minor axis of each tablet has a length of 2×b, is equal to the length 2×b minor �si inner surface of the shell, net Assembly clearance j formed between the pellets and the sheath, while the difference between half the length of the major axis of the truncated pills and a half length of the major axis of the shell (s-a) is much much more than the value specified for Assembly clearance.

For the purposes of the present invention, the expression "significantly" much more than the Assembly clearance, denotes the amount is so big, than the Assembly gap that the volumes of the internal cavities can be located with provision of swelling of the fuel without any peripheral interaction with the shell.

To implement technical solutions in accordance with the invention, the inventor has attempted to identify the mechanical processes occurring in the case of uncontrolled mechanical interaction between the tablet and the shell, in other words, in situations in which short-term mechanical interaction occurs outside the ability of the shell to withstand deformation.

Such situations arise, for example, when the reactor power increases to a level higher than the previously existing operating capacity or operating mode, in which the temperature of nuclear fuel does not activate the mechanisms of redistribution, in other words, it activates mechanisms autocompensation his own deformacije activates them only slightly.

In such situations, the existing fuel rods with a circular cross section is shown very strong mechanical interaction between the pellets and the sheath. In a rigid solid cylindrical tablets in such situations, there is a temperature gradient that decreases from the center towards the periphery; in other words, the cold periphery of the tablet has a radial stiffness that creates a certain kind of ring limiting stiffness. In addition, because the tablet only adapts quite a bit by itself, elasticity in the radial direction is not shown. Therefore, in such situations, the sheath has an annular limiting stiffness, called membrane rigidity created at the expense of the greater part of radial deformation of the fuel pellet. In other words, the annular restriction occurs in the radial direction of the interaction. Tablet, thus, allows only one possible direction of relaxation, namely the axial or longitudinal direction, which is provided by the local creep of the fuel in the direction of the depressions, formed for this purpose at the ends of each tablet.

The inventor also came to the conclusion that if thermomechanical characteristics of the fuel rod must be improved in the situation of face-to-face� strong mechanical interaction between the tablet and the shell, it should make the following decisions:

- to reduce the stiffness of the shell by changing the way of its annular restriction in the case of a circular cross-section to give the shell oval profile. Mechanical radial interaction between the tablet and the shell should not be axisymmetric. Thus, the original must be formed in an oval cross-section of the shell with a possible mechanical contact between the pellets and the sheath in the direction of small diameter (minor axis), and between the pellets and the casing must be formed in the gap so as to provide the movement, in other words, the extension of nuclear fuel in the direction of the large diameter (major axis),

accordingly, to reduce the stiffness of the cold periphery of the tablet by performing pills oval. This means that under the action of stresses on the pill, which has an oval shape, the interacting surface can be localized by performing their orthogonal only to the small diameter,

- to create not assymmetry temperature gradient in a tablet, having a temperature gradient to a greater extent, similar to the temperature gradient plate is cooled from two sides; not axisymmetric temperature gradient in the fuel can reduce the annular rigidity of the periphery of the cold table�current with a circular cross section, currently in use, through the creation of more hot sites at the ends of the major axis of the oval pill. This thermal effect contributes to the reduction of rigidity of oval profile, which tablet you may have along the minor axis,

- to create a greater volume of voids in the cross section (shell), so that the fuel, which swells and expands in diameter to redistribute themselves at the expense of fluidity without creating any tension or interaction between pellets and cladding. This redistribution at the expense of fluidity only possible if these voids are located in the vicinity of the hottest areas of the tablet, and reaction forces applied to the tablet during the interaction tablets with sheath, act on these very hot areas,

- to maintain the mechanical stability of the cross-section of the rod, to which is applied the pressure of the external cooler. In conditions of very strong mechanical interaction tablets and shell the resulting stiffness caused by the ovalization of the cross section must be sufficient to maintain the geometry of the cross section in a stable condition.

The inventor also proposes to do is carry the cross-section of the fuel rod elliptical shape for improved�of its thermo-mechanical characteristics in situations of mechanical interaction between the fuel pellets and the sheath.

In addition, the inventor has attempted to understand the other process occurring in the fuel elements during normal operation of the reactors that use these elements.

In existing reactors, such as water-water power reactors, fuel rod type elements formed from the pellets of nuclear fuel, having a round cylindrical shape, arranged separately to each other and placed inside the shell in the form of a tube having a length greater length of a column of tablets so as to leave the ends of column volumes to the extension necessary to limit the gradual increase of the pressure in the column of fuel elements under the action of gaseous fission products of nuclear fuel.

The transfer of heat between the tablets of nuclear fuel and the coolant occurs in the radial direction through thermal resistance created by the radial Assembly gap formed between the pills and the shell is filled with gas at the beginning of the period of operation of the fuel element, and the thickness of the material of the shell.

The control of this thermal resistance for the life of heating element ensures that the allowable limits fuel temperature will not be exceeded. The inventor therefore believes that the design of the new fuel�about rod must be taken into account the following factors:

the transfer of heat through the radial gas seal, calibrated at the beginning of the work period,

availability of the volumes formed in the transverse direction with respect to the direction of heat transfer.

Regular fuel elements of the plate type can compensate for the deformation caused by the impact of nuclear fuel, through the "plasticity" of their shells with very low mechanical strain in the shell, supporting at the same time, the transfer of heat in the warp direction. The inventor therefore believes that fuel cells must be made very narrow, in other words, you need to have in cross section a large ratio of length to thickness (width) to ensure that they can compensate for the deformation created by nuclear fuel in the thickness direction at very low values of stress in the shell.

Accordingly, the inventor came to the conclusion that the fuel element is made with a cross section in the shape of an ellipse, in accordance with the invention, should preferably be inherent in the three above-mentioned main features, in other words, it needs to have:

elliptical cross section in which the major axis has length 2×a, the minor axis is of length 2×b, the degree of splyusnutoi section is equal to the ratio a/b;

the tablet form should also be elliptical, forming in the Assembly radial clearance between the pellets and calibrated membrane, comparable to the gap existing in a typical fuel rods with a circular cross-section;

the presence of free volumes at the ends of the major axis of the tablet obtained by truncating the specified axis.

The inventor has found, thus, the solution disclosed in the present invention, namely tablets with elliptical cross section, a truncated along their major axis, separately of one another inside the shell is elliptical in shape with a radial clearance formed during Assembly along not truncated part tablets, cameras and expansion of gaseous fission products in the truncated ends of the tablets.

The result achieved at the expense of this new core cross section is desired improvement of thermo-mechanical characteristics in conditions of very strong mechanical interaction between the pellets and the sheath, due to the fact that:

such engagement is limited by the mechanical contact areas of the tablet and the shell, orthogonal to the minor axis of the cross section, which allows the shell to compensate for the deformation created by the pill, by reducing its elleptical form and, thus, within the thickness of the shell are then�Ko bending stresses localized at its the end portions along the major axis is 2×a;

the temperature gradient in the tablet allows greater flexibility mechanical properties of the tablets in the course of interactions;

the combination of the generally elliptical shape of the tablet and the presence of significant gas seals at the ends of its major axis generates a heat mainly in the direction of the minor axis when the hot core tablet that runs along the major axis and cold peripheral parts, limited sections that are in contact with the shell. Mechanical rigidity, which shows the tablet in the interaction in the direction of its minor axis, will be substantially reduced due to the almost complete absence of arch effect created by cold peripheral areas of the pill;

local resistance to the transfer of heat between the pills and the shell in the truncated ends of the tablet, in other words, along the major axes, increases the temperature of the surface area of the tablets in this area. Thus, in mechanical cooperation with the shell fuel tablet is compressed mainly along its small diameter, the presence of the hot zone to the surface at the ends of the major axis indicates that it may be deformed by creep mainly along the major axis. This �degree of freedom when you extrude due to creep in the direction transverse to limit the voids allows the tablet to compensate for the volume increase due to the creep mainly along this direction, minimizing appropriately deformation created by mechanical interaction tablets with sheath along its minor axis.

Specialists in the art will undertake efforts to ensure the geometric stability of the elliptical cross-section of the rod under the action of pressure forces applied to the fluid from an external source acting during normal operation of the reactor, in which the stiffness parameters associated with the fuel pill in order to prevent the flattening of the cross section.

These parameters may be the following:

the degree of splyusnutoi cross-section (ratio between the major and minor axes), which changes thermal characteristics of the tablet and, therefore, its stiffness to compress along the minor axis,

the dimensions of the cavities are located at the ends along a large truncated axis "C" tablets, which determine the temperature and therefore the amount of deformation creep tablets along this direction (rigidity against an extrusion direction of the cavities, which determines to some extent the rigidity of the pill to compress along its minor axis).

Thus, the new geometry of the rod, proposed according to the present invention, attached to the cross section geometric stability, assuring�I control the temperature gradient and the heat transfer from the tablet during normal operation, while at the same time compensation of distortions created by the action of the pills on the shell in a situation of mechanical interaction due to selection of the degree of splyusnutoi section and due to the structure with a truncation of the tablet and, therefore, buckling of the tablet at its ends, which minimizes the mechanical stresses in the shell due to the distribution of the deformations created between the tablet and the shell and also due to the chosen approach to solving the problem, according to which the shell is subjected to bending stresses due to the oval cross-sectional shape.

Preferably the Assembly clearance "j" is placed in the shell of the tablets formed within the truncated length of the major axis "C" is less than or equal to 10% of the length of the major axis 2×a shell.

If the terminal corresponding to the invention, intended for water-cooled power reactor (WWER), the shell is preferably made of a zirconium alloy or alloy M5 (ZrNbO), and the fuel pellet is preferably made of a ceramic material, such as UO2, (U, Pu)O2or from mixtures of oxide of uranium and recycled plutonium oxides.

If the terminal corresponding to the invention, intended for use in gas cooled nuclear reactors on fast neutrons (GBAR), the shell is preferably made of �rostockogo refractory or semi-refractory metal material, such, for example, alloys based on vanadium or of plastic ceramic material, such as, for example, Ti3SiC2belonging to the class of MAX phases, and the fuel pellet is preferably made of ceramic materials such (U, PU) C, (U, PU)O2.

The claimed invention also relates to a fuel Assembly with nuclear fuel containing a large number of fuel rods, such as described above and installed in the spacer grid.

In addition, the invention relates to a membrane made of a material transparent to neutrons, extending in the longitudinal direction, and with elliptical cross-section perpendicular to this longitudinal direction.

In addition, the invention relates to a tablet of nuclear fuel, which is located in the longitudinal direction and has a generally truncated elliptical shape with the major axis of the truncated cross-section perpendicular to the longitudinal direction of the tablets.

The invention relates also to a method of manufacturing a tablet of nuclear fuel, having in the longitudinal direction of the height H and in General a truncated elliptical shape with the major axis of the truncated, having a length of 2×for a, and minor axis of length 2×b in the cross section perpendicular to the longitudinal direction, wherein the method is carried out following this�s:

- preparation of powder of nuclear fuel at the stage of so-called pellets (tablet preparation),

- pressing powdered nuclear fuel along the contour of raw tablets implemented in a number of matrices of height H, made with a truncated elliptical cross-section with the length of the major axis is 2×C and the length of the minor axis is 2×b,

- sintering the molded tablet of nuclear fuel.

It should be noted that the term "raw tablet" means a tablet that was not prone to sintering.

Preferably, the ratio N/(2×(C) the height H to the length of the 2×with major axis equal to at least 1,2.

Thus, the new geometry of the fuel rod is described in accordance with the invention, also provides a possible alleged improvements in the manufacture of fuel rods. Truncated elliptical shape of the cross section of the fuel pellets means that the above two innovations in the manufacturing method below can be formulated differently:

- method of compressing tablets: a new form of tablets means that the axis line may be located along the direction of the minor axis of the elliptical cross-section (rather than the location of the axis of extrusion along the axis of the cylinder, as is known for tablets with a circular cross section). This new compression method can both�providing better control over the uniformity of the pressing density and, therefore, the geometry of the sintered tablets,

- exclusion of sollipulli carried out to fit the diameter of the tablet: a new elliptical shape of the cross section of the rod implies that the shell is forced into contact with the surface tablet (orthogonal minor axis) due to the action of external pressure as soon as the coolant temperature in the reactor increases.

Therefore, thermal performance of the tablet does not depend on initial Assembly clearance between the pellets and the sheath. Thus, unlike the situation in the prior art, there is no need for adjustment of the geometric dimensions of the tablets, since the deviation of the size obtained by sintering, become valid (in particular, due to the discussed above for a more perfect way of pressing).

The present invention relates also to method of stacking the fuel pellets in the shell from transparent to neutrons of material in such a way as to produce a fuel rod, in which the fuel pellet, obtained directly after sintering by using the above-described method of manufacture, is placed inside the shell having a generally elliptical shape, wherein the length of the minor axis of the inner (elliptic) surface of the shell is equal to 2×b and is equal to the length 2×b, minor axis of tablets plus the amount of Assembly gap, the difference between half the length of the major axis of the truncated pills and a half length of the major axis of the shell (s-a) much more than the value of Assembly clearance j.

Brief description of the drawings

Other advantages and characteristic features of the invention will become clearer from the following detailed description of the fuel rod according to the invention with reference to Fig.1 and Fig.1A, the contents of which are disclosed below.

Fig.1 shows a fuel rod according to the invention, a view in partial longitudinal section;

Fig.1A shows a fuel rod shown in Fig.1, a view in transverse section;

Fig.2 shows a fuel rod according to the invention, a perspective view;

Fig.3 shows a shell according to the invention, designed for placement of a column of tablets, one of which is shown in Fig.2, a perspective view.

The implementation of the invention

For a clearer understanding of the description of the longitudinal axis, along which are located the tablet 6, the shell 2 and the core 1 formed of these elements, marked the position XX/.

It should be noted that

the dimensions a and b are the inner dimensions of the elliptical shell 2,

the dimensions a and b are the external dimensions of the elliptical shell 2,

the size and/guide /fit uncut tablets 6,

size 2×with is the length of the major axis of 6 tablets of nuclear fuel, truncated in accordance with the invention.

Fig.1 shows a fuel rod 1 according to the invention, in a configuration ready for use in a nuclear reactor, in other words, being in vertical position, with tablets 6 placed closer to the bottom, as noted below. The rod 1 comprises a shell 2 made of a zirconium alloy, sealed at each of its ends with the upper plug 3 and the lower plug 4.

Inside the shell is essentially divided into two sections, one of which 5 is in the upper part, forming a chamber for expansion of the gas, and the other section houses is able to divide the column of tablets with 6 nuclear fuel, each of which is located in the longitudinal direction XX/rod 1.

In shown in the figure column each tablet 6 has approximately the same height H.

In the expansion chamber 5 is placed a helical compression spring 7, the lower end of which rests on a pillar of the tablets 6, and the other end of the spring rests on the top cap 3.

The spring 7 holds the post of tablets in place in position along the longitudinal axis XX/perceives longitudinal thickening of 6 pills in the process and, in �wow, prevents buckling of the cross section of the shell having an oval shape.

In other words, the spring prevents excessive ovalization section of the shell.

Fig.1 shows the correct (undeformed) cross-section of the rod 1.

Shell 2 in accordance with the invention has a uniform thickness around the perimeter and a generally elliptical shape. More specifically, the inner surface 200 of the shell 2 elliptical has a large axis of length 2×a and minor axis of length 2×b.

Tablet 6 nuclear fuel has an elliptical shape, truncated at each end of the major axis of the shell. In other words, the tablet has 6 large truncated axis of length 2×C and a minor axis of length 2×b/.

It should be noted that the size determines the distance from the plane of truncation of the tablets 6 to its center.

Constant radial Assembly clearance j between the tablet 6 and the shell 2 is formed on the elliptical sides of the tablet, in other words, the entire length of the 2×with pills. In other words, immediately after manufacture and prior to use as nuclear fuel in a nuclear reactor each fuel tablet 6 has a truncated elliptical cross section in which the length of the b/half of the minor axis approximately equal to the half length b of the minor axis of the inner surface 200 of the shell 2 net Assembly clearance j.

p> Free volume or expansion of the cavity 60 are located on two ends of the major axis of the truncated tablets 6, in other words, between the truncated end 61 of the tablet 6 and the inner surface 200 of the shell 2.

Thus, the cross-section of the fuel rod 1 can be expressed on the basis of the geometrical parameters of the tablets 6 as follows.

The ovality coefficient or the coefficient of splyusnutoi section pills "a//b/"where and/=a-j,

The coefficient truncation tablets "/a/".

The inventor believes that the coefficient of splyusnutoi "a//b/"must be equal to at least 1,5 to achieve a satisfactory thermal characteristics typical fuel plate described in the international application WO 2007/017503.

Possible to use the fuel rod 1 with an elliptical cross-section in accordance with the invention in two types of nuclear reactors, operating with the coolant, the cooling of the active zone, which is supported at a higher pressure than the pressure in the fuel elements.

The first scheduled use of the rod is in its use with operating parameters implemented in water-water power reactors (VVER).

The rod, therefore, is a priori can be made�Yong from the same composite materials that use for typical designs of existing fuel elements, such as rods with a circular cross section, such currently known; namely, zirconium alloys or alloy M5 (ZrNbO) for shell and ceramic materials or a mixture of oxide of uranium and recycled plutonium oxides for fuel pellets.

The second intended use of the fuel rod is in its use for gas cooled fast reactors, i.e. in conditions in which the cladding temperature is high, is within the range of from 300°C to 900°C, and is high intensity of fast neutron flux. Composite composite materials used for the manufacture of the rod, can be heat or half heat-resistant metals, for example for the manufacture of membranes used alloys based on vanadium ceramic or plastic materials, such Ti3SiC2belonging to the class of MAX phases, and for fuel pellets - ceramic materials, such (U, PU) C or (U, PU) O2.

Below is described one specific embodiment of the rod with elliptical cross-section, corresponding to the present invention. In this embodiment, the terminal 1 is designed to match the operating parameters of a typical pressurized water energeticheskogo (VVER).

Below compares the geometric parameters, materials and operating parameters of a typical pressurized water reactor.

The dimensions of the rod, made with the famous circular cross-sections:

Sheath: outer diameter D dia = 9.5 mm, inner diameter D EXT = 8,36 mm.

Fuel pellets: diameter = 8,2 mm.

Materials:

The shell is made of alloy M5, fuel pellets from UO2.

Performance parameters:

The temperature on the outer surface of the shell T = 342°C, a pressure of the fluid P = 155 bar, power per unit volume of nuclear fuel = 320 W/cm3.

The rate of burnup of nuclear fuel = 60000 MW day/tonne.

Based on these comparative data for the terminal with the known circular cross section, the author of an invention offered for the new elliptical rod according to the invention the following geometric dimensions:

Section tablets such as tablets with ordinary round cross-section,

The ovality coefficient and//b/= 1,8,

The degree of truncation equal to C/a/= 0,9, in this case the rod 1 is characterized by the following dimensions and/b/with:

and/= 5,61 mm, b/= 3,115 mm/= 5,05 mm,

shell thickness 0.57 mm equal to the thickness of the shell with the usual round cross-section,

the radial Assembly gap is equal to the radial Assembly gap between the tablets and the shell in the rod with �the usual round cross-section, in which the value of j ≈ 0.08 mm. the size of the gap j, formed in the modular structure of the rod between the tablets 6 and the shell 2 in accordance with the invention, measured along the shorter axis b of the ellipse, while the dimensions of the elliptical cross-section of the shell as follows:

large internal axis 2×a = 5,69 mm

minor internal axis 2×b = 3,195 mm

large external axis 2×A = 6,26 mm

small external axis 2×In = 3,765 mm.

In relation to the selected geometry comparison of conventional rod with a round cross-section, designed for water-cooled power reactor (WWER), the total cross section of the rod 1 with an elliptical cross-section in accordance with the invention is increased about 4.4%, and occupied by the fuel cross-sectional area is approximately 92.5% of the cross-section of the shell.

Thus, the total cavity j, 60, formed the initial radial Assembly clearance j between the 6 pills and sheath and planes of truncation of the ends 61 of 6 tablets (cavity 60 between cut side surfaces 61 and the inner surface of the shell 2), is approximately 7,47% of the internal cross-section of the shell equal to P×a×b.

In the manufacture of the shell 2 with an elliptical cross-section, there is no special technological problems.

For the manufacture of tablets 6 can also be used different pressing operation. The coefficient SP�znalosti and ,/b,described in the present invention, equal to 1.8 in the above dimensions, means that it becomes possible to carry out the pressing of each tablet orthogonal, in other words, along the direction of the minor axis and,its elliptical cross-section or, in other words, on its lateral surface bounded by a height H, and not along the axis XX/cylindrical surface, as is done currently for rods with a circular cross-section.

The elliptical shape of the shell also means that the tablet after sintering can be placed in the shell. The inventor believes that the pressing of the fuel pellet at its side surface with a height H should lead to a smaller variation of thickness of the sintered tablets due to the greater uniformity of density within the tablet after pressing.

As noted above, in the process of VVER elliptical shape of the shell would mean that due to the location of the coolant under pressure will be the contact between the outer surfaces of tablets and shell (except surfaces facing the terminal cavities 60), in other words, the entire length of the 2×S.

Even at the beginning of the service life of thermal characteristics of the tablets 6 are no longer dependent on initial Assembly clearance between Tadla�kami 6 and the shell 2.

The analysis of thermal and thermo-mechanical characteristics of the rod with elliptical cross-section corresponding to the invention, the operation conditions of the reactor VVER, chosen as the base of the reactor was carried out using computer modeling using finite elements (CAST3M).

The modeling was based on the assumption of the constancy of power given off by the nuclear fuel during the period of operation, changing the physical properties of materials M5 shell and material UO2 nuclear fuel depending on the temperature, the viscoelastic properties of the shell material and fuel (thermal creep and creep caused by radiation), swelling of materials under neutron irradiation and the intensity of release of gaseous fission products produced by nuclear fuel, about 6% (which is a typical value set for rods with a circular cross section, operating at a speed of fuel burnup of 60,000 MW day/tonne).

The results obtained for the core with a specified rate of fuel burnup is equal to 60000 MW day/tonne indicate the following:

- good temperature control during the operation period; since the beginning of the power generation radial clearance j between the tablets 6 and the shell 2 is eliminated, and the maximum tempera�Hooray nuclear fuel is changed from the temperature of the beginning of life, equal to 683°C to a final temperature of this period of 904°C.

This change is due to the deterioration of thermal conductivity of the fuel under irradiation and the presence of gaseous fission products emitted by the fuel, which reduce the heat transfer coefficient between the tablets 6 and the shell 2.

Due to the elliptical cross-sectional shape, since the size of the tablet along the direction of heat transfer (the minor axis) is less than the diameter of a round tablet with the same cross-sectional area, the maximum temperature inside the fuel is less than in a conventional rod with a round cross-section;

- good General thermomechanical characteristics in the cross section of the tablet of nuclear fuel. This provides control over the deformations of the cross section, since the fluidity of the elliptical cross-section of the tablet of nuclear fuel is determined by the surface temperature, achieved through a thermal resistance created by the cavities 60 formed in the truncated ends 61 pills.

In the early period of operation local temperature increases (on the side surfaces 61) and at 136°C higher than the temperature of heat-transfer areas (on sites 62) in contact with the shell.

At the end of life, the increase of the local temperature (between cut ends of sections 61 and 62) is 220°C.

This heat of rawnow�this, which controls the mechanical stability of the cross section, obtained by optimization of geometrical parameters of the cross section, namely ovality ratio a/b and the degree of truncation of the C/a. Obviously, these parameters depend on each application and its optimization depends on the operating parameters of each tablet of nuclear fuel and the mechanical properties of the component materials, in particular from temperature creep and patterns of behavior under irradiation.

Good thermomechanical characteristics also lead to a good control of the internal pressure in the rod created by the gaseous fission products, a dedicated nuclear fuel.

The presence of cavities 60 in the truncated ends 61 of the pill forms the additional expansion chamber, which are absent in the bar having a conventional circular cross-section.

Finally, good thermomechanical characteristics create mechanical interaction between the pill 6 and the shell 2, which curves the shell.

Created localized bending stresses in the end portions 200 of the shell facing surfaces 61 truncation tablet of nuclear fuel.

The fluidity of the material of the shell 2 limits during the operation of these mechanical stresses values constituting less than 100 MPa.

Consequently, the shell of things�stvu is under the action of bending stresses, acting on her oval; it is not exposed annular limiting stiffness, as can occur with the shell bar having a conventional circular cross-section.

Deformation in the cross section of the fuel pellet 6 is compensated mainly due to the extrusion of the fuel due to the fluidity in the direction of terminal cavities 60 under the influence of the stiffness of the oval shape of a truncated elliptical cross-section of the tablet, which thus counteract the deformation of expansion and swelling.

Other improvements and modifications can be made without going beyond the scope of the invention.

For use in water-water power reactors (VVER), currently in operation, could be used conventionally used materials, namely, the zirconium alloy for the shell 2 and UO2for fuel pellets 6 or mixture on the basis of depleted uranium oxides and recycled plutonium oxides, also called Moss. Characteristics of the core can be optimized by controlling the nature of the change in creep of the sheath materials and fuel rod with elliptical cross-section in accordance with the invention.

For use in the gas-cooled fast reactors (BR) use of plastically deformable membrane is desired �ri its manufacture of some plastic metal and ceramic materials, the above.

1. Fuel rod (1) located in the longitudinal direction (XX'), containing a large number of tablets (6) nuclear fuel stacked on one another, and the shell (2) made of a material transparent to neutrons, surrounding the column of tablets, wherein in a cross section perpendicular to the longitudinal direction (XX'):
- shell has an essentially elliptical shape, and the inner surface (200) has a greater axis of length 2×a and minor axis of length 2×b,
- each tablet (6) nuclear fuel is made elliptical in shape, truncated at the ends of the major axis of the shell, wherein the minor axis of each tablet has a length of 2×b' equal to the length of 2×b of the minor axis of the inner surface of the shell minus the value of Assembly clearance j formed between the pellets and the sheath, while the difference between half the length of the major axis of the truncated pills and half the length of the major axis of the shell (s-a) is much more than the Assembly clearance j.

2. The fuel rod according to claim 1, in which Assembly the clearance j between the pellets and the sheath along the length of the major axis of the truncated 2×with less than or equal to 10% of the length of the major axis is 2×and the inner surface of the shell.

3. The fuel rod of claim 1 or 2 for the water-water energy reactor (WWER), whose shell is made of an alloy of zirconium or alloy M5 (ZrNbO), and tablets nuclear t�fuel made from ceramic materials, such as UO2, (U, Pu)O2or from mixtures of oxide of uranium and recycled plutonium oxides.

4. The fuel rod according to claim 1 or 2 for gas-cooled nuclear reactors on fast neutrons (BR), in which the shell is made of heat-resistant refractory or semi-refractory metal material, such, for example, alloys based on vanadium, or of plastic ceramic material, such as, for example, Ti3SiC2belonging to the class of MAX phases, and fuel pellets made of ceramic materials, such (U, PU) C, (U, PU)O2.

5. A fuel Assembly with nuclear fuel containing a large number of fuel rods according to any one of claims.1-4, placed together in the lattice spacing.

6. Tablet (6) nuclear fuel, located in the longitudinal direction (XX') and in cross section perpendicular to the longitudinal direction (XX') having essentially a truncated elliptical shape with the major axis of the truncated.

7. A method of manufacturing tablets (6) nuclear fuel height H in the longitudinal direction (XX') and in cross section perpendicular to the longitudinal direction (XX') having essentially a truncated elliptical shape with the major axis of the truncated length of 2×for a and minor axis of length 2×b', when this method is carried out in the following stages:
- preparation of powder of nuclear fuel�VA at the stage of so-called pellets,
- pressing powdered nuclear fuel along the contour of raw tablets implemented in a number of matrices of height H, made with a truncated elliptical cross-section with the length of the major axis is 2×C and the length of the minor axis is 2×b',
- sintering the molded fuel pellets.

8. A method of manufacturing according to claim 7, in which the ratio N/(2×C) between the height H and length 2×with major axis equal to at least 1,2.

9. A method of laying tablets (6) nuclear fuel in a shell (2) made of a material transparent to neutrons, so to get a fuel rod, in which tablets of nuclear fuel after sintering, made directly using the method of manufacturing according to claim 7 or 8, placed inside the shell, having an essentially elliptical shape, wherein the minor axis of the inner surface of the shell has a length of 2×b, net Assembly clearance j is equal to the length 2×b' of the minor axis of tablets, the difference half the length of the major axis of the truncated pills and half the length of the major axis of the shell (s-a) is much more than the Assembly clearance j.



 

Same patents:

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FIELD: physics.

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The invention relates to nuclear engineering, in particular to designs of fuel elements for fast reactors with liquid metal coolant

Nuclear reactor // 2236047
The invention relates to the field of nuclear energy and can be used in high temperature nuclear reactors with helium coolant

FIELD: nuclear power engineering; fuel rods for water-moderated water-cooled reactors.

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SUBSTANCE: proposed method involves sequential fluid-bed deposition of coating layers onto fuel microspheres. First low-density pyrocarbon layer is deposited by pyrolysis of acetylene and argon mixture of 50 volume percent concentration at temperature of 1450 °C. 85 - 95 % of second layer is deposited from high-density pyrocarbon by pyrolysis of acetylene and argon mixture of 40.0 - 43,0 volume percent concentration, and of propylene and argon mixture of 30.0 - 27.0 volume percent concentration at temperature of 1300 °C; 5 - 15 % of coating is deposited by pyrolysis of propylene and argon mixture of 5.0 - 10.0 volume percent concentration doped with 0.5 - 1. 5 volume percent of methyl trichlorosilane. Third layer of silicon carbide is deposited by pyrolysis of methyl trichlorosilane and argon mixture of 2.5 - 3.0 volume percent concentration in hydrogen-argon mixture at temperature of 1500 °C. Upon deposition this layer is treated with hydrogen at temperature of 1750 -1800 °C for 20 - 30 minutes. 90 - 95 % of fourth layer is deposited by pyrolysis of acetylene and argon mixture of 40.0 - 43.0 volume percent concentration and of argon and propylene mixture of 30.0 - 27.0 volume percent concentration at temperature of 1300 °C. Upon deposition of 90 - 95 % of fourth-layer pyrocarbon coating thickness 5 - 10 % of coating is deposited by pyrolysis of propylene and hydrogen mixture of 3.0 - 5.0 volume percent concentration.

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FIELD: physics.

SUBSTANCE: invention is designed for increasing operation parametres and life cycle of active zone of a reactor due to the maximum fuel burnup, improved reliability and operation safety of maintenance of nuclear power stations. Fuel element can of a fast reactor with liquid metal heat carrier includes a metal tube of vanadium alloy with titanium, chrome and unavoidable impurities. External and internal tube surfaces are coated with stainless ferrite steel. Vanadium alloy components are taken at a given ratio. In particular, titanium to chrome ratio lies within 2.2 to 1.8 range. Between vanadium alloy and stainless ferrite steel a 6-8 mcm thick transition layer of solid solution of vanadium alloy with stainless ferrite steel is formed.

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EFFECT: reduced probability of formation of voids, transfer of pressure onto the moulded workpiece becomes more uniform, increased output of the suitable products.

FIELD: machine building.

SUBSTANCE: procedure consists in immersion of tubular casing of fuel rod into water electrolytic medium containing particles of iron oxide and in covering it at least partially with layer of iron oxide. Also, particles of iron oxide are produced by anode oxidising iron containing working electrode.

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SUBSTANCE: invention relates to nuclear reactor fuel rods. Fuel rod cladding has an elliptical cross-section. Each nuclear fuel pellet along the longer axis of the cladding has a truncated elliptical shape, and the shorter axis of the pellet has the same length as the shorter axis of the cladding, minus the fitting gap j, wherein the difference in length of the longer axis of the cladding and the truncated longer axis of the pellet is much greater than said fitting gap j. The invention also relates to a method of making nuclear fuel pellets and a method of packing said pellets, which enables to form a fuel rod.

EFFECT: lower probability of deformation of fuel rod cladding and release of fission products into the coolant.

9 cl, 3 dwg

FIELD: engines.

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8 cl, 3 dwg

FIELD: metallurgy.

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4 cl, 4 dwg, 1 tbl, 4 ex

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