JPS6350416B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION Field of the Invention This invention relates to improvements in nuclear fuel elements for nuclear reactor cores, and more particularly to improved nuclear fuel elements having a cladding vessel with a metallurgical gradient in the cross-section of the vessel wall.

BACKGROUND OF THE INVENTION A variety of nuclear reactors are currently being designed, constructed, and operated in which nuclear fuel is contained in fuel elements having various geometries, such as plates, tubes, or rods. Usually the fuel material is surrounded by a corrosion-resistant, non-reactive, thermally conductive container or cladding. A plurality of fuel elements are assembled in a grid at fixed mutual spacing within a coolant flow channel or region to form fuel assemblies, and a sufficient number of fuel assemblies are assembled to form a fission chain reaction assembly. or to form a reactor core capable of self-sustaining nuclear fission reactions. On the other hand, the reactor core is housed within a reactor vessel, and coolant is passed through the reactor vessel.

Cladding, or cladding, serves several purposes, two of which are primarily
first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or moderator, if present, or both coolant and moderator, if both are present; and second, to prevent radioactive fission production. The objective is to prevent substances, some of which are gases, from being released from the fuel into the coolant or moderator or both, if both are present. The coating material is stainless steel,
Common materials include aluminum and its alloys, zirconium and its alloys, niobium, and certain magnesium alloys. If the cladding ruptures, i.e. loses leak-tightness, the coolant or moderator and related systems become radioactive long-lived products.
There is a risk of contamination to the extent that it will interfere with plant operations.

Important requirements for materials used in nuclear reactor construction include low thermal neutron absorption, corrosion resistance, high ductility and mechanical strength. Zirconium-based alloys fully meet these requirements and are therefore widely used for this type of purpose, with Zircaloy 2 and Zircaloy 4 being two examples of important commercially available alloys useful for such applications. It will be done.
These alloys exhibit corrosion under normal boiling water reactor operating conditions, resulting in thick oxide spalling from the channels and thick oxide on the fuel rods. Spalling of oxide flakes can, in some cases, generate high radiation fields in the vicinity of the control rod mechanisms where such flakes collect, and the presence of thick oxide layers reduces heat transfer efficiency. This results in localized heating of the fuel cradling.

It would be desirable to improve the hot water and steam corrosion resistance of zirconium-based alloys without sacrificing other properties of tubes made from such alloys. The difference in corrosion resistance between the inner and outer peripheries of the cladding tube can be obtained by creating a composition gradient. For example, corrosion resistance on one side can be enhanced by plating or otherwise forming a composite structure. That method is expensive. It is desirable to form a tube of uniform composition with excellent corrosion resistance.

SUMMARY OF THE INVENTION According to the invention, a particularly effective nuclear fuel element cladding tube for a nuclear reactor is provided, which cladding tube is characterized in that the cross section of the tube wall has a metallurgical gradient.

The metallurgical gradient has a relatively less corrosion resistant metallurgical condition near the inner circumference of the tube and a relatively higher corrosion resistant metallurgical condition near the outer circumference.

According to the invention, a method is also provided for generating such a metallurgical gradient in the cross-section of the wall of a zirconium alloy cladding tube, which initially has a metallurgical state of relatively low corrosion resistance. The exterior of the tube is uniformly heated to at least the high alpha range for a time sufficient to transform the exterior of the tube to a relatively corrosion-resistant metallurgical state. At the same time as heating the outside of the tube,
The interior of the tube is maintained at a temperature at which substantially no metallurgical changes occur and substantially no oxidation of the inner surface occurs.

After heating the outside of the tube, the tube is cooled quickly enough to maintain a relatively corrosion-resistant metallurgical state at the outer periphery. Preferably, heating and cooling of the exterior of the tube is carried out in the presence of an inert fluid to prevent the formation of oxides on the exterior surface of the tube.

DETAILED DESCRIPTION OF THE INVENTION The above and other objects and advantages of the present invention will be better understood from the following detailed description taken in conjunction with the accompanying drawings.

The principal field of application of the principles of the invention is in the fabrication of nuclear fuel assemblies such as the one shown in partially cut-away side view in FIG. The fuel assembly 10 is a typical fuel assembly for a boiling water nuclear reactor, and includes a tubular flow channel 11 with a substantially square cross section, a lifting handle 12 at the upper end of the flow channel 11, and a lifting handle 12 at the lower end of the flow channel 11. A normal nozzle is provided (note that the nosepiece is not shown since the lower part of the fuel assembly 10 has been omitted from the drawing). The upper end of the channel 11 is open with an outlet 13, and the lower end of the nosepiece is provided with a coolant flow opening. An array of fuel elements or fuel rods 14 containing nuclear fuel material is surrounded by channel 11 and supported within the channel by a top plate 15 and a bottom plate (not shown). Typically, coolant enters the channel through an opening at the lower end of the nosepiece and flows upwardly along the fuel element 14, partially evaporated in the case of boiling water reactors and non-evaporated in the case of pressurized water reactors. It exits from the upper outlet 13 at high temperature.

A nuclear fuel element or rod 14 has its ends covered with a cladding tube 1.
It is sealed with an end plug 18 welded to 7 and has a stud 19 to facilitate attachment to the fuel rod assembly. A void space or plenum 20 is provided at one end of the fuel element to permit longitudinal expansion of the fuel material and to store gases released from the fuel material. Nuclear fuel material restraining means 24 in the form of a helical spring member are located in the space 20 to restrain axial movement of the nuclear fuel material, in particular during handling and transport of the fuel element.

The fuel element is designed to provide good thermal contact between the cladding and the fuel material, minimize neutron absorption, and resist warping and vibrations occasionally caused by high velocity coolant flow.

A nuclear fuel element or rod 14 constructed in accordance with the principles of the present invention is shown in partial section in FIG. Fuel element 14 comprises a core or column of nuclear fuel material 16, shown here as a plurality of fuel pellets of fissile material and/or fuel parent material, disposed within a structural cladding or vessel 17. In some cases, the fuel pellets may be of various shapes, such as cylindrical pellets or spheres, and in other cases, different fuel forms may be used, such as granular fuel. The physical form of the fuel is not important to the invention. A variety of nuclear fuel materials can be used as the nuclear fuel, including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. The preferred fuel is uranium dioxide or a mixture of uranium dioxide and plutonium dioxide.

Turning now to FIG. 2, the nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a cladding tube 17. During use in a nuclear reactor, cladding 17 surrounds the fissile core, leaving a gap 23 between the core and the cladding. FIG. 2 is not drawn to scale; the distance of the gap 23 and the wall thickness of the cladding tube 17 are exaggerated for illustrative purposes. The cladding tube is made of zirconium alloy. Preferably, the cladding is made of Zircaloy 2 or Zircaloy 4. Zircaloy-2 is approximately 1.5% by weight
Contains tin, 0.12% iron, 0.09% chromium, 0.005% nickel and 0.1% oxygen. Zircaloy 4 is substantially similar to Zircaloy 2 except that it contains almost no nickel and about 0.2% iron.

Zirconium alloys typically have an intermetallic grain phase. As mentioned above, suitable zirconium alloys contain tin, iron and chromium, and may also contain nickel. Suitable zirconium alloys contain the intermetallic compound Zr(Cr,Fe) ₂ and may also contain _Zr2 (Ni,Fe) in the form of granular precipitates.

Pure zirconium has two different crystal lattice structures or phases, each stable over different temperature ranges.
That is, it has α and β phases. Besides, zirconium alloys, such as Zircaloy 2 and Zircaloy 4, have a stable mixture of two crystal lattice structures in a third intermediate temperature range.

Here, the term "α crystal structure" or "α phase"
means a hexagonally packed crystal lattice structure that is stable at relatively low temperatures. The temperature range in which the α phase is stable is referred to as the "α range."

The term "β crystal structure" or "β phase" also refers to a body-centered cubic crystal lattice structure that is stable at relatively high temperatures. The temperature range in which the β phase is stable is the “β range”
It is called.

In pure zirconium, the alpha crystal structure is stable up to about 860°C (1580°). A phase change occurs at approximately this temperature, and a stable β crystal structure occurs at temperatures above approximately 860°C (1580°C). Zirconium alloys have α near the phase change temperature of pure zirconium.
and β crystal structure has a stable temperature range. The particular temperature range over which this mixture is stable depends on the particular alloy. For example, Zircaloy 2
α at a temperature of 810℃ (1490〓) to about 971℃ (1780〓)
Indicates a stable mixture of +β crystal structures.

The cladding or vessel has a metallurgical structure across its wall that varies from a relatively less corrosion resistant metallurgical condition at the inner circumferential surface 24 to a relatively more corrosion resistant metallurgical condition at the outer circumferential surface 26. It has a slope.

Typically, the intermetallic phase in sections of the tube with a relatively less corrosion-resistant metallurgical state is approximately uniform in shape, and the intermetallic phase in the relatively more corrosion-resistant sections is at least partially for example, into two dimensional arrays. However, it has been determined that the metallurgical gradient has no detectable difference in the shape of the precipitate phase between relatively corrosion-resistant and relatively corrosion-resistant metallurgical states. Ta.

The present invention further provides a method of manufacturing a zirconium alloy cladding tube having a metallurgical gradient as described above. The method begins by forming a cladding tube, which is typically in a metallurgical state with relatively low corrosion resistance.
This is the most stable state of the alpha phase, and is typical of tubes that have not yet been heated to a temperature at which relatively high corrosion resistance properties are exhibited.

The exterior of the tube is heated to a temperature sufficient to transform that portion into a metallurgical state that is highly corrosion resistant. It is currently preferred to heat the outside of the tube to a high alpha range or a mixed alpha+beta range.

There is a “high α” temperature range in which the α phase is stable,
It was confirmed that zirconium alloy heated to this range transforms into a metallurgical state with relatively high corrosion resistance. Zirconium alloys that are heated to the high alpha range and then quenched quickly enough to maintain a relatively corrosion-resistant metallurgical state exhibit high corrosion resistance. For the preferred zirconium alloys, this range ranges from approximately 704°C (1300°C) to when the alloy is in the α phase to mixed α
up to the temperature at which it undergoes a phase change to the +β phase.

It is preferable to set the upper limit of the above temperature range to the upper limit at which the mixed α+β phase is stable. Although heating to the β range provides the desired relatively corrosion-resistant metallurgy, it requires more energy and time to heat and cool the tube and reduces the ductility of the tube. High alpha range or mixed alpha+beta instead of beta range
It was confirmed that heat treatment within this range has far less effect on the ductility of the zirconium alloy. However, the present invention as described above allows the outer portion of the tube to be
This can be done by heating to temperatures ranging up to the β range of 982°C (1800°C) or higher.

By flowing a coolant through the tube while heating the exterior of the tube, the interior surface is maintained at a sufficiently low temperature to maintain a relatively corrosion-resistant metallurgical state and to prevent oxide formation. The use of coolant is especially necessary for thin-walled tubes, eg, tubes with wall thicknesses of about 1 inch or less. This is because zirconium alloys have a high thermal conductivity and practically cannot add and remove heat to the outside of the tube at a rate that does not heat the inside portion of the tube to undesirable temperatures.

Preferably, the internal temperature does not exceed about 427°C (800°C). Above about 427°C (800°), oxide formation occurs on the inner surface. More preferably, the internal temperature does not exceed about 99°C (210°C). The inner temperature is about 210
If it is below, no substantial steam will be generated in the pipes even if water at normal pressure is used as the coolant.

After heating the outer surface of the tube sufficiently to transform the alloy in that portion to a relatively more corrosion-resistant metallurgical state, preventing substantial transformation of the tube to a relatively less corrosion-resistant metallurgical state. Cool quickly enough. This results in a tube having a modified high α crystal structure or a modified mixed α+β crystal structure on the outside of the tube.

Heating a section of the zirconium alloy tube to the high alpha range for a period sufficient to transform the section to a relatively more corrosion-resistant metallurgical state, and then transforming it back to a relatively less corrosion-resistant metallurgical state. A portion is described herein as having a "transformed high alpha" crystal structure when cooled to substantially prevent .

Similarly, a section of zirconium alloy tube that has been heated to a temperature sufficient to cause a phase change from an alpha phase to a mixed alpha+beta phase and then cooled is described herein as having a "transformed alpha+beta" crystal structure. do.

It is also preferred that the heat treatment is carried out in the presence of an inert fluid to prevent the formation of oxides on the outer surface of the tube.

FIG. 3 shows an apparatus for heat treating cladding tubes according to the present invention. A preferred method of achieving the metallurgical gradient as described above includes uniformly heating the outer circumferential portion of the cladding tube 31 to at least the high alpha range, for preferred zirconium alloys to about 704°C (1300°C) or higher. During this time, the inner peripheral surface of the tube is maintained at a temperature below about 427°C (800°), preferably below about 99°C (210°). The cladding tube 31 is mechanically coupled through an induction coil 33 to tube advancement means, such as rollers 32 . The tube is advanced at a constant linear travel speed to achieve uniform heat distribution to the tube. The preferred speed is approximately 10-76 cm/min (4-30 cm/min).
inches/minute) or higher. The speed of travel is adjusted to provide sufficient residence time in the induction coil so that the tube receives an amount of heat that transforms the outer circumferential portion of the tube to a relatively corrosion-resistant metallurgical state.

Heat is generated by an induction coil. The induction coil is typically energized with alternating current in the frequency range of 3000-50000 Hz. The power used should be sufficient to generate the required amount of heat. The area within the induction coil of the tube is heated to a temperature sufficient to cause a metallurgical transformation to a more corrosion resistant state at its outer surface.

The extent or depth of transformation to a relatively corrosion-resistant metallurgical state depends on the time and temperature of the tube within the induction coil. The temperature depends on the input power to the coil. A radiant pyrometer 34 senses the external temperature of the heated region of the tube and controls the temperature to a selected value by adjusting the input power to the coil 33 via a standard electronic feedback control system 36 and matching station 37; limit.

Oxide formation on the outer surface of the cladding tube is minimized by blanketing the hot area within the induction coil and the surrounding environment with an inert fluid such as helium.
That is, the hot region is located within a quartz tube 38 that is joined to an inlet box 39 and an outlet box 41.
An inert gas is introduced into the outlet box 41 through a pipe 42 and is forced into the hot area inside the quartz tube by the flow resistance to the exit gas stream by the inlet gland 43 and outlet gland 44 at the locations where the tube enters and exits the device. Maintain a slight positive gas pressure.

While heating the exterior of the tube, cooling fluid is passed through the flexible line 46 into the tube to cool the interior surface of the tube. The preferred coolant is water. The flowing coolant in contact with the inner surface of the tube maintains the inner portion of the tube in a relatively less corrosion resistant metallurgical state. The inner surface of the tube is
Because it is kept at a low temperature, it does not undergo chemical reactions or oxidation during heat treatment. Once the heated region of the tube passes the induction coil and there is no more energy to enter and heat that section of the tube, the water flowing inside the tube rapidly cools the outer section of the tube by heat conduction through the tube wall. Cooling is sufficiently rapid to maintain the relatively corrosion-resistant metallurgical state created by the induction coil.

FIG. 4 shows a section of the tube passing through the induction coil. As the tube 48 is advanced, a hot region 47 of the tube 48 is created within the induction coil 49 . The water 51 flowing inside the tube cools the inner part 52 of the tube 48 and remains unaffected by heat. As the tube advances, the induction coil 49
The outer portion 53 of the tube, which had been heated by the tube, is cooled by conduction, and the heat is transferred to the water flowing inside the tube. This creates a metallurgical gradient in which the outer portion 53 of the tube has been transformed by the heat treatment to a relatively corrosion-resistant metallurgical state, and the inner portion 52 is unaffected by the heat treatment. Although the inner and outer portions 52 and 53 of the tube are shown in the figures as two separate layers, metallurgically they actually have progressive corrosion resistance.

Although heating the tube by induction is the preferred heating method, there are many other rapid heating methods that can achieve the desired results. For example, by passing a current from one end of the tube to the other, the entire tube can be heated to a desired temperature range, ie at least to the high alpha range. This can be done by attaching a ring or circular electrical contact to each end of the tube. The current is adjusted so that the current resistance of the tube is sufficient to heat the tube. The cooling liquid flowing inside the tube sufficiently cools the inner surface of the tube to prevent the formation of oxides, and also cools the entire tube sufficiently quickly to maintain the relatively corrosion-resistant metallurgical state generated by heat treatment after the current is cut off. hold. This heat treatment method is also preferably carried out in an inert atmosphere to prevent the formation of oxides on the outer surface of the tube.

The desired metallurgical gradient can also be achieved by heating the tube in a furnace while flowing a cooling liquid inside the tube. However, this method makes it more difficult to prevent oxide formation on the outer surface of the tube.

The heat treatment according to the invention can be carried out at any stage of tube forming. The method is particularly useful for treating final products, which can be heat treated without undergoing significant deformation. However, it is currently preferred to carry out a heat treatment prior to the final cold forming process. This gives the tube a greater wall thickness, which requires relatively less energy to heat the outer surface of the tube to the desired temperature. This is because heat removal by conduction through the tube wall is slower than in tubes with smaller wall thicknesses. The cold tube forming process performed after the heat treatment proportionally reduces the metallurgical properties of the cladding wall, but does not significantly reduce the beneficial effects achieved with this heat treatment.

The more corrosion resistant metallurgical state of the tube is more resistant to oxidation in hydrothermal and steam environments than the less corrosion resistant metallurgical state. The cladding obtained by using the above method has a relatively corrosion-resistant metallurgical condition on the exterior of the tube, which is typically the only part in contact with steam and hot water, and on the entire tube. The fibers maintain desirable mechanical properties, namely high ductility.

Example 1 A Zircaloy 2 tube shell having a diameter of 2.5 inches, a length of 6 feet, and a wall thickness of 0.430 inches was placed within a four-turn induction coil.
Fill the inside of the tube shell with water for approximately 19 minutes (5 gallons/
minutes). The induction coil was energized with a 200 kilowatt, 3000 Hz power supply through a suitable matching transformer.

The temperature outside the tube shell inside the induction coil was raised to approximately 899℃ (1650℃) in approximately 22 seconds. The temperature outside the tube shell was maintained at about 899°C (1650°C) for about 8 seconds. Next, the power was cut off, and the tube shell was allowed to cool to about 204°C (400°C) in about 21 seconds while water was flowing inside the tube.

Next, the heat-treated tube shell is transferred to a Pilger tube making machine for 30 minutes.
The tube was reduced to a final cladding tube having an outer diameter of about 0.483 inches and a wall thickness of about 0.032 inches. For each forming (diameter reduction) pass, the tube was annealed at approximately 621°C (1150°) for approximately 2 hours.

The tube shell was then subjected to a steam corrosion test at approximately 500°C (932°C) for approximately 24 hours. The corrosion weight gain on the heat-treated section of the tube was about 1/4 of the weight gain exhibited by the non-heat-treated section of the tube.

Since the inner part of the tube unit is kept at a relatively low temperature during the heat treatment of the present invention, and thus there is no significant change in this area, the present invention provides an inner barrier against fission products or reactor It can be applied to tube units having various internal parts, such as coatings or linings, which serve other functions as fuel containers within the tube unit. Tubes for nuclear fuel Kratzdig to which the present invention is applicable include tubes with a zirconium metal barrier as disclosed in U.S. Pat. No.
No. 3969186, No. 3925151, No. 4022662, No.
4045288 and 4316771.

FIG. 5 shows a barrier-lined tube heat treated in accordance with the present invention. In this example, the cladding or container 17 for the fuel 16 is provided with an outer circumferential region 26 having a relatively corrosion-resistant metallurgical condition, as well as a liner 54 bonded to the inner circumference or inner surface 24. has been done. Liner 54 may be made of zirconium or other metals previously used as barriers by those skilled in the art.

[Brief explanation of drawings]

FIG. 1 is a partially cross-sectional side view of a nuclear fuel assembly having a nuclear fuel element constructed in accordance with the principles of the present invention, and FIG. 2 is a nuclear fuel element having a cladding tube manufactured in accordance with the principles of the present invention. FIG. 3 is a perspective view of an apparatus for heat treating cladding according to the principles of the present invention, and FIG. 4 shows a portion of the cladding in the heating region of the apparatus shown in FIG. 5 is a cross-sectional view of a nuclear fuel element having a barrier lined cladding made in accordance with the principles of the present invention; FIG. 10... fuel element, 16... fuel substance, 17...
...cladding tube, 24...inner circumference, 26...outer circumference, 31...
... Cladding tube, 33 ... Induction coil, 47 ... High heat area, 48 ... Cladding tube, 49 ... Induction coil, 51
...water, 52...inner part, 53...outer part,
54... Raina.

Claims (1)

[Claims] 1. The inner periphery of the zirconium alloy tube is in a metallurgical state with relatively low corrosion resistance, and the outer periphery is in a metallurgical state with relatively high corrosion resistance, with a metallurgical gradient. A method for producing a zirconium alloy tube includes forming the zirconium alloy tube in a metallurgical state with relatively low corrosion resistance, and subjecting the outer periphery of the zirconium alloy tube to a high-α temperature range or mixing ( α + β) temperature range to convert said outer periphery to a relatively corrosion-resistant metallurgical state while simultaneously cooling said inner periphery while substantially metallurgically changing the inner surface. and rapidly cooling the zirconium alloy tube to maintain a metallurgical state with relatively high corrosion resistance in the outer periphery. 2. The method of claim 1, wherein the intermetallic particle phase is substantially uniformly distributed in metallurgical conditions with relatively low corrosion resistance. 3. The method according to claim 1, wherein the intermetallic particle phase is distributed in at least a partially segregated manner in a metallurgical state of relatively high corrosion resistance. 4. The method according to claim 1, wherein the outer peripheral portion is heated to 704°C (1300°C) or higher. 5. The method of claim 1, wherein the outer peripheral portion of the zirconium alloy tube is heated by passing the zirconium alloy tube through an induction coil energized with alternating current. 6. The method of claim 1, wherein the outer peripheral portion of the zirconium alloy tube is heated by passing a substantially uniform electrical current through the zirconium alloy tube. 7. The method of claim 1, wherein while heating the outer peripheral portion of the zirconium alloy tube, the inner peripheral portion of the zirconium alloy tube is cooled by continuously passing a coolant through the interior of the zirconium alloy tube. Method. 8. The method according to claim 1, wherein after heating, the zirconium alloy tube is cooled by passing a coolant through the inside of the zirconium alloy tube. 9. The method according to claim 1, wherein heating and cooling are performed while preventing the formation of oxides on the outer peripheral surface of the zirconium alloy tube. 10. The method according to claim 9, wherein the formation of oxides on the outer circumferential surface of the zirconium alloy tube is prevented by heating and cooling the outer peripheral portion of the zirconium alloy tube in the presence of an inert fluid. . 11 The temperature of the inner peripheral surface of the zirconium alloy tube is
The method according to claim 1, wherein the temperature does not exceed 427°C (800°C). 12 The temperature of the inner peripheral surface of the zirconium alloy tube is
The method according to claim 1, wherein the temperature does not exceed 99°C (210°C).