WO2022223504A1

WO2022223504A1 - Nuclear fuel pellet comprising a heat-conducting metal or metal alloy insert having solid disks or disks with a hole along the central axis

Info

Publication number: WO2022223504A1
Application number: PCT/EP2022/060218
Authority: WO
Inventors: Vincent Fabreguettes; Bernard Valentin
Original assignee: Commissariat A L'energie Atomique Et Aux Energies Alternatives
Priority date: 2021-04-19
Filing date: 2022-04-19
Publication date: 2022-10-27
Also published as: FR3122026B1; FR3122026A1

Abstract

The invention relates to a nuclear fuel pellet (6) comprising: - a straight cylinder (60) of fissile material having a central axis (X) of which the length and diameter define the length (H) and the diameter (Øpellet), respectively, of the pellet, the straight cylinder being bored at its center; - a heat-conducting metal or a metal alloy insert (8), the insert extending along the length of the pellet and comprising at least one solid or centrally bored disk (80), the disk being centered in relation to the central axis, the ratio R between the diameter of the disk (Ødisk) and the diameter of the pellet (Øpellet) being less than or equal to 0.8.

Description

Description Title:

NUCLEAR FUEL PELLET INCORPORATING A THERMAL CONDUCTIVE METALLIC OR METALLIC ALLOY INSERT WITH SOLID DISCS OR HOLE ACCORDING TO THE CENTRAL AXIS

Technical area

The present invention relates to the field of fuel elements for Pressurized Water Reactors or Pressurized Water (PWR) also referred to by the English acronym P WR (Pressurized Water Reactor).

More precisely, it is in the field of ceramic-type fuels made of uranium oxide (UO ₂ ).

The invention essentially aims to improve the thermal properties of these fuels, both in nominal operation and in an Accidental Loss of Primary Coolant (APRP) situation.

By "nuclear reactors", throughout the application, is understood the usual meaning of the term to date, namely plants for the production of energy from nuclear fission reactions using fuel elements in which the fissions which release the calorific power, the latter being extracted from the elements by heat exchange with a coolant which ensures their cooling.

By "nuclear fuel rod", throughout the application, we understand the official meaning defined for example, in the dictionary of nuclear sciences and techniques, namely a narrow tube of small diameter, closed at both ends, constituting from the core of a nuclear reactor and containing fissile material. Thus, a "nuclear fuel pin" whose use favors the name is a nuclear fuel rod within the meaning of the present invention.

Prior technique

Nuclear reactors that use fission energy to produce heat can be classified into several different categories depending on their characteristics: the form of final energy produced (electricity, heat, etc.), the type of neutron flux ( fast neutrons or thermalized neutrons), the cooling fluid used (liquid metal, water, etc.), the physical state of the cooling fluid (solid, liquid or gaseous), the level of coolant pressure (e.g. atmospheric for Boiling Water Reactors and high for PWRs) ...

One sector is currently very largely industrially dominant in the world, that of Pressurized Water Reactors (PWR).

A PWR essentially comprises two main components: the primary circuit which constitutes the nuclear part of the reactor, and the secondary circuit which constitutes the non-nuclear part of it. For example, you can refer to the principle diagram under the link https://fr/wikipedia.org/wiki/R%C3%A9acteur_%C3%a0_eau_pressuris%C3%A9e

The primary circuit essentially comprises a vessel assembly with the internal structures and the fuel assemblies.

The secondary circuit essentially comprises non-nuclear equipment to produce electricity (piping, steam generators, turbine, etc.). Steam generators are heat exchangers that transfer the heat from the core to the secondary fluid, i.e. water in a PWR, which vaporizes then expands in a turbine which, together with the alternator, forms the heat generation system. 'electric energy.

To return to the primary circuit, this incorporates the key part, called the reactor core. The core groups together all the blends, of which there are 157 in a PWR of the French sector. A schematic view of a PWR fuel assembly with its control cluster comprising rods of neutron-absorbing material is shown in figure 4 under the link: http://www.cea. ff /Documents: monographs: combustibles- nucl%C3%A9aires_r%C3%A9acteurs-eau.pdf

The core is itself placed in a vessel which constitutes the second containment barrier for the fuel and its fission waste (called Fission Products or FP).

The core is cooled by a heat transfer fluid, water at a pressure of 150 bars, which also acts as a moderator (slowing down the neutrons), in order to promote the fission reaction.

This core also includes internal structures performing specific functions, such as maintaining the assemblies, channeling the heat transfer fluid, etc., which are not detailed here. The fuel assemblies make it possible to produce energy by taking advantage of the nuclear fission reaction of a fuel composed for a small part of heavy fissile nuclei - the isotope 235 of uranium - as well as the isotope 238 of uranium which is only fertile (it will produce plutonium 239 by neutron capture).

This fuel is introduced into a cylinder of circular cross-section closed at each of these two ends by a plug. This cylinder is called pencil and described more precisely below. The rod is sealed and constitutes the first containment barrier.

The pencils are grouped in bundles and arranged in a square pitch network. For the families of installation of the French sector called "bearings" of 900 MWe and 1300MWe, the number is 264 rods to which are added 25 other tubes, which then composes an assembly which also incorporates many structural elements (plate of foot, retaining grids, an instrumentation tube, 24 guide tubes where 23 control rods slide, head plate, retaining spring, spider for handling the control rods) as shown in figure 4 of the aforementioned .pdf document .

In Figure 1, there is shown a nuclear fuel rod 1 according to the state of the art which is shown in its configuration of use in a PWR nuclear reactor, that is to say in a vertical position with the pellets 6 towards the lower part as specified below.

The rod 1 consists of a sheath 2 conventionally made of Zircaloy-4 (Zr4) closed at each of its ends by a respectively upper 3 and lower 4 plug which is welded thereto. This waterproof pencil is filled with helium at 25 bars when cold to partially counterbalance the effect of the external pressure of 150 bars of the heat transfer fluid.

The interior of the sheath is essentially divided into two compartments, one of which 5 in the upper part, between the top of the fissile column and the upper plug 3, constitutes a gas expansion chamber and the other houses the column fissile formed by the stack of nuclear fuel pellets 6 which each extend along the longitudinal direction XX' of the rod 1.

The expansion chamber is a free volume intended to receive FPs in gaseous form, usually called Fission Gas (GdF). In the stack shown, each patch 6 has substantially the same length or height H.

A helical compression spring 7, generally made of Inconel®, is housed in the expansion chamber 5 with its lower end resting against the upper face of the highest pellet 6 of the stack of pellets and its other end resting against the top cap 3.

In addition to maintaining the stack of pellets 6 along the longitudinal axis XX' and the "absorption" over time of the longitudinal swelling of the pellets 6, the other function of this spring 7 is to prevent the buckling of the section of the sheath on its mode of ovalization. In other words, it must prevent the extreme ovalization of the sheath section.

The primary function of a fuel rod is to produce and then transmit the heat produced by the fission reactions within the fuel.

To date, a fuel pellet as implemented in a PWR reactor consists of uranium oxide U02 enriched in U235 to about 5%, the remainder being fertile U238.

Each pellet releases, by nuclear fission, energy in the form of heat and which varies over time depending on the wear of the fuel but also on the variation in the altitude of the control rods.

The power thus dissipated is also a function of the position of the pellet in the rod, the position of the rod in the assembly and the position of the assembly in the core.

This power is evacuated to the cold source of the primary circuit (the primary cooling water) by encountering a certain number of thermal obstacles which can be summarized as follows:

- a strong gradient between the center and the periphery of the pellet induced by the low thermal conductivity of U02;

- a radial thermal gradient between pellet and rod sheath. Indeed, the seal between pellet and sheath, entirely gaseous (helium) at the start of irradiation, becomes completely filled at the start of the second cycle. However, the roughness of the pellet allows the discontinuous presence of gas which is no longer simply helium but also includes GdF: thus, the contact between pellet and sheath is never perfect and therefore creates, by its thermal resistance, a radial thermal gradient;

- the radial thermal conduction through the sheath;

- the resistance due to the radial convective exchange between the external face of the sheath and the heat transfer fluid.

The removal of heat in the nominal operating regime and in incidental and accidental regimes is governed by the heat conduction equation (or Fourier equation) with energy dissipation for the fuel pellet to which is added the Newton's equation which models the convective exchanges by a coefficient of heat exchange between the sheath and the heat transfer fluid.

These equations apply in incidental and accidental situations, as long as the geometry of the rod (including pellets) remains intact and the coolant has not evaporated.

Numerical simulations of a Loss of Primary Coolant (LOCA) type accident have demonstrated the interest, from the point of view of the temporal thermal consequences of this accident, of having at the rated and therefore initial operating conditions of this accident, a fuel as cold as possible at heart.

The main objective is therefore, in terms of safety, to lower the thermal of the rod and more particularly of the fuel pellet.

Figure 2, taken from publication [1], gives the order of magnitude of the temperatures in a fuel pellet in nominal operating conditions.

There are different ways to improve the power removal function produced by a U02 fuel pellet or fuel pellet stack. Since some characteristics can hardly be modified, such as the primary fluid, the sheath material, the most open avenue for innovation is that of improving the thermal conductivity of U02.

This path can be divided into two categories which have been explored.

The first category concerns the improvement of the thermal conductivity without adding another phase within the U02 fuel. This first category can itself be subdivided into three sub-categories as follows:

- a modification of the shape of the rod, in order to increase the exchange surface between hot source and cold source. This modification is localized by definition at the level of each rod of a fuel assembly. This solution has the major disadvantages of requiring a new design of the assembly and of having a possible impact on the manufacturability of the rod and the performance of the assembly;

- a change in the nature of the fuel, passing from a combustible oxide to a better heat-conducting fuel, for example a metallic fuel. This modification is localized by definition at the level of each fuel pellet. This solution has the major drawbacks of inducing a different behavior in the reactor, of requiring research and development to develop/qualify this new fuel, of developing new manufacturing processes and of having to redefine the downstream fuel cycle;

- a combination of the two previous sub-categories which is a concept referred to as LIGHTBRIDGE: see in particular [2] This concept consists of a generally helical rod whose sheath contains a metallic fuel, UZr. In addition to the drawbacks explained for the two previous sub-categories, this solution has the major drawbacks of greatly reducing the mass of fissile material, typically by around 25% and of requiring over-enrichment of the pellet in U5 fuel.

The second category concerns the improvement of the thermal conductivity with the addition of another phase within the U02 fuel. The addition is localized by definition at the level of each fuel pellet.

This second category can itself be subdivided into three sub-categories as follows:

- a dispersion of diamond, graphene... on a nanometer scale. This solution does not seem relevant to date from the point of view of its thermal behavior. The inventors believe that it is however difficult to decide on this solution due to a lack of information and knowledge. This solution has the major drawbacks of being only on the scale of a laboratory and therefore, with the need for complete research and development to be carried out for the fuel, to analyze the manufacturability and then to put implement research and development, and finally to have any modeling tools;

- a homogeneous dispersion of a second metallic phase on a microscopic scale with the aim of reducing the thermal conductivity of the U02 pellet: see in particular [1] This solution has the major drawbacks of inducing a manufacturing difficulty and requiring the finalization and then the validation of existing modeling tools;

- a heterogeneous dispersion of a second metallic phase on a macroscopic scale with the aim of promoting the heat flow from the U02 pellet to the cold source. This second metallic phase is characterized by a metallic insert within the fissile material of the pellet.

Such a metal insert is described under two distinct designs in the publication [3].

One of these designs, shown in figure la of this publication [3] consists of a set of wafers distributed over the length of the pad. This design effectively improves the thermal of the fuel pellet in the center. However, the analysis made by the applicant's experts on this design shows that if, indeed, the choice of the structural characteristics of the wafers makes it possible to achieve a significant reduction in the temperature of the core pellet, on the other hand this choice potentially generates several disadvantages that can have very annoying consequences on the behavior of the pellet, or even completely call into question the relevance of the solution.

The main potentially negative consequences are:

- a hypothetical guarantee of the thermal function under irradiation and of the mechanical resistance to irradiation of the wafers given their low thickness;

- mechanical strength of the sheath not guaranteed, with high and local temperature gradients thereon, typically from 100 to 300°C/mm, which are due to the contact of each wafer with the sheath;

- the lack of certainty of not aggravating the phenomenon of pad sheath interaction with stress corrosion (IPG-SCC);

- a level of corrosion likely to be present on the sheath during irradiation;

- manufacturability of the metal wafers in a ceramic fuel not demonstrated with in particular the possibility of mounting the wafers and the absence of guarantee of an optimal mounting of the fuel column. There is a need to improve the thermal design of nuclear fuel pellets, more particularly based on U02 oxides, in order to reduce their core temperature as much as possible, in nominal operating conditions for a Pressurized Water Reactor core ( REP) while respecting the following constraints:

- minimization of the quantity of second phase provided in a tablet;

- absence of degradation or at least limited degradation of neutron behavior;

- manufacturability as easy as possible;

- assembly of a fuel column as simple as possible;

- Absence of the aforementioned drawbacks of the design shown in Figure la of this publication [3].

The object of the invention is to meet this need at least in part.

Disclosure of Invention

To do this, the invention relates, in one of its aspects, to a nuclear fuel pellet, comprising:

- a straight cylinder of fissile material with a central axis (X), the length and diameter of which respectively define the length (H) and the diameter of the pellet; the cylinder

right being pierced in its center;

- a metal or thermally conductive metal alloy insert, the insert extending over the length of the pad and comprising at least one solid or drilled disc in its center, the disc being centered on the central axis, the ratio R between the diameter of the disc and the

diameter of the pellet being less than or equal to 0.8, preferably between 0.7

and 0.8.

According to an advantageous embodiment, the pellet comprises four solid discs regularly spaced along the central axis (X), the solid rod connecting the solid discs to each other.

According to a first variant, each disk is flat.

According to a second variant, each disc consists of a central portion and an annular portion around the central portion, the annular portion being two-sided spherical of the same diameter which are opposed forming a biconcave ring of greater thickness at the central axis (X) and at the periphery of the pellet.

Advantageously, the material of the insert is chosen from a zirconium alloy, in particular Zircaloy-4 (Zr4), chromium (Cr), Molybdenum (Mo).

Advantageously again, the fissile material of the right cylinder is chosen from uranium(IV) oxide (U02), mixed oxide (U, Pu)02 or a mixed mixture based on uranium oxide and reprocessed plutonium oxides.

Preferably, the mass percentage of the insert is between 5 and 10%.

The invention also relates to a nuclear fuel rod extending in a longitudinal direction (XX) comprising:

- a plurality of nuclear pellets as described above, stacked on top of each other;

- a neutron-transparent material sheath surrounding the stack of pellets.

Advantageously, the sheath is made of zirconium alloy, in particular Zircaloy-4 (Zr4), or M5® alloy (ZrNbO).

The invention also relates to a nuclear fuel assembly comprising a plurality of fuel rods as above and arranged together in a network.

The invention also relates to the use of a nuclear fuel pellet as described above or a fuel rod as described above in a pressurized water reactor (PWR).

Thus, the invention essentially consists of a nuclear fuel pellet which incorporates, within a straight cylinder of fissile material, a metal insert with solid disc(s) and solid rod along the central axis, as as second metallic phase.

With an insert with solid discs and solid rod according to the invention, the core temperature of a pellet is not the lowest among all the possible solutions.

On the other hand, the insert allows a significant drop in the core temperature, while not generating the disadvantages caused by the solutions of the prior art, in particular the design shown in figure lb of this publication [3]. The solution according to the invention is a perfect compromise for simultaneously obtaining several performances on a complex system constituted by a nuclear fuel pellet.

The advantages of a fuel pellet according to the invention are numerous, including:

- a large drop in maximum temperature compared to that of a pellet according to the state of the art;

- strong limitation of local thermal gradients on the sheath of a rod which houses a pellet according to the invention;

- simple manufacturability;

- maintenance of a usual assembly process in a nuclear fuel rod;

- an increase in fission gas retention;

- a limitation or even elimination of the risk of IPG-CSC because, by design, the height of a pattern composed of U02 is less than the diameter of the fuel pellet (U02 + metal insert).

Other advantages and characteristics of the invention will emerge better on reading the detailed description of examples of implementation of the invention given by way of illustration and not limitation with reference to the following figures.

Brief description of the drawings

[Fig 1] Figure 1 is a schematic view in longitudinal section of a nuclear fuel rod according to the state of the art, as implemented in a PWR nuclear reactor.

[Fig 2] figure 2 illustrates in the form of a curve the temperatures in a nuclear fuel pellet according to the state of the art, in nominal operating conditions.

[Fig 3 A] Figure 3 A is a schematic perspective view of a basic pattern M of a nuclear fuel rod whose sheath houses a fuel pellet with a metal insert with discs drilled in their center according to the invention. [Fig 3B] Figure 3B repeats Figure 3A but without the presence of the sheath, ie only the fuel pellet with a metal insert with solid discs and solid rod according to the invention.

[Fig 3C] Figure 3C reproduces Figure 3B but without the presence of the right cylinder of fissile material.

[Fig 4] Figure 4 is a view from the numerical simulation showing the temperature fields for the configuration of Figure 3 A.

[Fig 5] Figure 5 is a view from the numerical simulation showing the temperature fields for the configuration of Figure 3C.

[Fig 6A] Figure 6A is a schematic perspective view of a fuel pellet without the metal insert according to a first variant of the invention.

[Fig 6B] Figure 6B is a schematic perspective view of the metal insert according to a first variant of the invention.

[FIG 6C] FIG. 6C is a view resulting from the digital simulation showing the temperature field for a pellet configuration according to FIG. 6A with an insert according to FIG. 6B.

[Fig 7] Figure 7 is a view from the numerical simulation showing the temperature fields for the configuration of Figure 6A.

[Fig 8] Figure 8 is a view from the numerical simulation showing the temperature fields for the configuration of Figure 6C.

[Fig 9 A] Figure 9 A is a schematic perspective view of a second fuel pellet variant without the metal insert according to the invention.

[Fig 9B] Figure 9B shows is a schematic perspective view of a second variant of a pierced disc of the metal insert according to the invention.

[FIG 9C] FIG. 9C is a view resulting from the numerical simulation showing the temperature field for a patch configuration according to FIG. 9A with an insert according to FIG. 9B.

[Fig 10A] Figure 10A is a schematic perspective view of a third fuel pellet variant without the metal insert according to the invention. [Fig 10B] Figure 10B shows is a schematic perspective view of a third variant of a pierced disc of the metal insert according to the invention.

[FIG 10C] FIG. 10C is a view resulting from the digital simulation showing the temperature field for a pad configuration according to FIG. 10A with an insert according to FIG. 10B.

detailed description

For the sake of clarity, the same element according to the state of the art and according to the invention is designated by the same numerical reference.

Figures 1 and 2 have already been commented on in the preamble. They will therefore not be detailed below.

In the examples below, the fuel pellets according to the invention and the cladding are created using the CAD software marketed under the name "SolidWorks" and then modeled, by a mesh and 3D thermal conduction calculations, using its "Professional Simulation" module in the 2019 version. In all of the figures 3 A to 10C, even if the various elements are not represented exactly to scale, the insert with solid discs and rod solid according to the invention is illustrated as it is, i.e. on a macroscopic scale, in a cylinder of fissile material.

There is shown in FIG. 3A a basic pattern M of a nuclear fuel rod 1 with a sheath 2 housing a fuel pellet 6 according to the invention. As illustrated in FIG. 3B, the pellet 6 according to the invention with a central axis (X) comprises a straight cylinder 60 of oxide combustible material releasing thermal power and a metal insert 8.

The metal insert 8 comprises a plurality of discs 80 pierced in their center by a central hole 81, of circular section, of equal thickness Ep and regularly spaced from each other along the length H of the pellet.

The right cylinder 60 is machined with a central hole 61 and at least one circular recess 62 in order to accommodate a solid disc 80 therein. FIG. 3C shows the metal insert 8 on its own, with its dimensions represented, namely the diameter of the solid discs 80, their constant thickness Ep, the diameter

of the central hole 81.

The thermal gain of a pellet 6 with metal insert 8 according to the invention is evaluated by comparison with the maximum temperature of the oxide fuel (denoted below T _{max comb} ) of a standard cylindrical pellet of circular section of a PWR in service in the 900 MWe series of the French sector.

Since the calculation on a single pattern increases the maximum temperature of the fuel, a stack of four patterns M is necessary to converge towards a correct limit value. This stack of four patterns therefore comprises a number of four discs 80 pierced in their center at regular intervals.

The standard cylindrical patch as shown in FIG. 1 has the same dimensions as a patch 6 according to the invention, ie the same diameter and the same length H.

The power of the basic pattern M in accordance with the invention is 373 W, which corresponds to standard operating conditions at nominal speed.

The percentage of the metal insert 8 in a pellet 6 is 10%. Its material is the same as that of sheath 2, namely zircaloy 4 (Zr4).

The geometric dimensions of a patch 6 according to FIGS. 3A to 3C are as follows:

- pellet diameter equal to 8.36 mm;

- diameter of a full disc 80 equal to 6.57 mm (hence a report

equal to 0.79;

- thickness of a solid disc 80 (Ep) equal to 0.4 mm;

- diameter of a central hole 81 equal to 1.5 mm.

With this geometry, the drilled discs 80 of the insert 8 are never in contact with the sheath 2, regardless of the irradiation history.

The objective of such a configuration is to limit the very local thermal gradients on the sheath 2, typically of the order of at least 150° C./mm if there was perfect contact between discs and sheath, and its consequences on the sheath 2.

The temperature T _{max comb} b of a standard fuel pellet is 1093°C.

The estimated gain on T _{max comb} in these conditions is approximately 226°C. FIGS. 4 and 5 show the thermal fields respectively for the base pattern M for four pellets 6, and the metal insert 8 with drilled discs 80 with central hole 81. The local temperature gradients on the sheath 2 are reduced by one factor of about 9, to be close to about 15 to 17°C/mm. In order to reduce the mass percentage of the metal insert 8 in a pellet 6, the inventors propose replacing the Zr4 with a more heat-conductive material.

A parametric study on potency shows that gains increase with potency as shown in Table 1 below.

It is specified that in this table 1, the powers retained for the calculations are those dissipated by a pellet of the fuel column and all the pellets are assumed to have the same power.

The power of 373 W is representative of the maximum power of a standard PWR chip in nominal operating conditions.

A power of 500 W is already a power in accident conditions. Table 1 below gives an estimate of the potential gains in temperature (δT _max equal to T _{max comb} pellet according to the state of the art - T _{max comb} of an insert 8 according to the invention) according to the power dissipated in the pellet and for a metal insert 8 occupying 10% of the volume of the pellet, the ratio R =

is equal to 0.79.

It is specified that in this table 1, the values δT _max 10% Cr _t , δT _max 10% Mo, δT _max 10% Zr, designate the gains respectively for an insert made of chromium, molybdenum, and zircaloy.

[Table 1]

Table 1

In order to reduce the mass percentage of the metal insert 8 in a pellet 6, the inventors propose replacing the Zr4 with a more heat-conductive material.

Chromium (Cr) and molybdenum (Mo), despite its neutron penalty, are retained for their better thermal conductivity.

Table 2 below summarizes the calculations to evaluate the influence of the choice of the material of the metal insert 8 on the gains in temperature of lU02, all the dimensions being kept with the same ratio=0.79. Only the thickness Ep of the discs

varies to get 7.5% and 5% metal.

[Table 2]

Table 2

The temperatures in table 2 are compared with that of a pellet according to the state of the art with 10% Zr4 (T _max UO2 = 866°C) to propose a target percentage range of a metal insert 8 other than the Zr4 in the pellet. The calculations in Table 2 reveal that:

- the choice of Cr or Mo makes it possible to reduce the quantity of the metal insert 8 in a range situated between 5.8% and 6% with Cr and 4.8% and 5% with Mo.

- in these ranges with low addition of metal: o the design of the pellet must be more accomplished and satisfy other thermal, thermomechanical and/or thermochemical criteria, o this makes it possible to limit the over-enrichment necessary to compensate for the reduction in mass of uranium, o the quantity of waste to be processed will be more limited. - the thermal fields are similar to those of the initial reference case above;

- by design (choice of ratio

0.8), the local temperature gradients on the sheath are low (< 5%/mm) or even negligible (< 0.5°C/mm). These are therefore no longer a discriminatory criterion.

Different geometric variants of the metal insert 8 can be envisaged, in order to optimize the dimensional structure parameters diameter/thickness of the discs, diameter of the central hole and the functional parameter maximum temperature of the fuel.

Table 3 below gives an estimate by calculation of the potential gains in temperature (δT _max equal to T _{max comb} pellet according to the state of the art - T _{max comb} of an insert 8 according to the invention) according to the power dissipated in the pad and for a metal insert 8 in Zr4 occupying 10% of the volume of the pad.

[Table 3]

Table 3

From this table 3, it emerges that the gains are close in the range 0.78<<0.8 with a difference of about 4°C. This small difference is explained by the fact that the dim

inution of (therefore of distance from the cold source) is compensated by increase of

in an area where the fuel is hottest. This range (from 0.78 to 0.8) is an optimum and gives a geometric margin for thermomechanical dimensioning.

There is no exploration beyond the report

> 0.8 because this increases the thermal gradients on the sheath, which is even more penalizing at high power and would remove margins in incidental or accidental situations involving an increase in power or temperature by an external cause.

A first geometric variant of the metal insert 8 consists in modifying its shape and thus, instead of a drilled disc 80 with central hole 81, to produce a solid disc 80.

This first variant is illustrated in FIGS. 6A to 6C.

Figures 7 and 8 show the thermal fields respectively for the basic pattern M for four pellets 6, and the metal insert 8 with solid discs 80. The calculations show that the maximum temperatures of the fuel are identical - except for rounding - to those with disc drilled in their center with a central hole 81, and this regardless of the choice of the nature of the disc material. Thus, here too The local temperature gradients on sheath 2 are reduced by a factor of around 9, to be close to around 15 to 17°C/mm. A second geometric variant of the metal insert 8 consists in modifying its shape and thus, instead of a flat drilled disc 80 with central hole 81, to produce an annular portion 80b with two spherical faces of the same diameter which are opposed forming a biconcave ring of greater thickness at the central axis (X) and at the periphery of the pellet. This second variant is illustrated in FIGS. 9A to 9C.

In the case retained, the maximum thickness of the disk is 1 mm at its center and its periphery. The calculations show that the maximum fuel temperatures are identical - to the nearest degree - to those with discs drilled in their center with a central hole 81, and this regardless of the choice of disc material. If this second variant does not present any gain in temperature, it makes it possible to further reduce the localized thermal gradients on the sheath by approximately -5°C/mm, which should contribute to further limiting the possible disadvantages of a pellet according to the state of the art. A third geometric variant of the metal insert 8 consists in modifying its shape and thus, instead of a flat drilled disc 80 with central hole 81, to produce an annular portion 80b with two spherical faces of the same diameter which are opposed forming a biconcave ring of greater thickness at the central axis (X) and at the periphery of the pellet. This third variant differs from the second variant in that the central hole 81 is removed.

This second variant is illustrated in FIGS. 10A to 10C.

In the case chosen, the maximum thickness of the disc is 1 mm at its center and its periphery. The calculations show that the maximum fuel temperatures are identical - to the nearest degree - to those with a disc drilled in their center without a central hole 81, and this regardless of the choice of disc material. If this third variant does not present any gain in temperature, it makes it possible to further reduce the localized thermal gradients on the sheath by approximately -5°C/mm, which should contribute to further limiting the possible disadvantages of a pellet according to the state of the art.

The invention is not limited to the examples which have just been described; it is in particular possible to combine together characteristics of the examples illustrated within variants not illustrated.

Other variants and improvements can be considered without departing from the scope of the invention.

List of cited references

[1]: Kim DJ. - Rhee Y.W. & al - “Fabrication of Micro-Cell J 02 -Mo with enhanced thermal conductivity”, JNM 462 (2015) 289-295.

[2]: Malone J. - Totemeier A. - Shapiro N. - Vaidyanathan S. - Lightbridge Corp - “Advanced Metallic Fuel for LWRs”, Nuclear Technology, 181:3437-442 Nov 2012.

[3]: Medvedev P.G. & Mariani R.D. - “Conductive inserts to reduce nuclear fuel”, JNM 531 (2020) 151966.

[4]: H. Bailly & al. - “The nuclear fuel of pressurized water reactors and fast neutron reactors” - Eyrolles 1996

Claims

1. Nuclear fuel pellet (6), comprising: a straight cylinder (60) of fissile material with a central axis (X) whose length and diameter respectively define the length (H) and the diameter

the pellet; the right cylinder being pierced in its center; an insert (8) made of metal or a thermally conductive metal alloy, the insert extending over the length of the pellet and comprising at least one disc (80) solid or drilled in its center, the disc being centered on the central axis , the ratio R between the diameter of the disc ( and the diameter of the pellet being less than or equal to 0.8, preferably

between 0.7 and 0.8.

2. Pellet (6) according to claim 1, comprising four solid discs evenly spaced along the central axis (X).

3. Pellet (6) according to one of claims 1 or 2, each disc being flat.

4. Disc (6) according to one of claims 1 or 2, each disc consisting of a central portion of an annular portion around the central portion, the annular portion (80b) having two spherical faces of the same diameter which are opposed forming a biconcave ring of greater thickness at the central axis (X) and at the periphery of the pellet.

5. Pellet (6) according to one of the preceding claims, the material of the insert being chosen from a zirconium alloy, in particular Zircaloy-4 (Zr4), chromium (Cr), Molybdenum (Mo).

6. Pellet (6) according to one of the preceding claims, the fissile material of the right cylinder being chosen from uranium(IV) oxide (UO ₂ ), mixed oxide (U, Pu)02 or a mixture mixed based on uranium oxide and reprocessed plutonium oxides.

7. Pellet (6) according to one of the preceding claims, the mass percentage of the insert being between 5 and 10%.

8. Nuclear fuel rod (1) extending in a longitudinal direction (XX') comprising: a plurality of nuclear pellets (6) according to one of the preceding claims, stacked on top of each other; a sheath (2) made of material transparent to neutrons surrounding the stack of pellets.

9. Pencil (1) according to claim 8, the sheath being made of zirconium alloy, in particular Zircaloy-4 (Zr4), or M5® alloy (ZrNbO).

10. Nuclear fuel assembly comprising a plurality of fuel rods according to any one of claims 8 or 9 and arranged together in a network.

11. Use of a nuclear fuel pellet (6) according to one of claims 1 to 7 or of a nuclear fuel rod (1) according to claim 8 or 9 in a pressurized water reactor (PWR).