WO2010142693A1

WO2010142693A1 - Method for determining the maximum operating values of a nuclear reactor to avoid damage by pellet-cladding interaction

Info

Publication number: WO2010142693A1
Application number: PCT/EP2010/058030
Authority: WO
Inventors: Christian Royere
Original assignee: Areva Np
Priority date: 2009-06-08
Filing date: 2010-06-08
Publication date: 2010-12-16
Also published as: FR2946454B1; FR2946454A1

Abstract

The present invention relates to a method for protecting against fuel cladding failure caused by pellet-cladding interaction in a pressurised-water reactor, characterised in that the method comprises steps which involve determining the maximum linear power density (Pl max) of the fuel rods and the drift of said linear power density over time (formula 1); determining a linear power density threshold (PL) according to said drift of said maximum linear power density (formula 1); comparing said maximum linear power density (Pl max) with said linear power density threshold (PL); triggering an alarm and/or a corrective action to stop the power augmentation if said maximum linear power density (Pl max) is no lower than said predetermined linear power density threshold (PL).

Description

METHOD FOR DETERMINING THE LIMITS OF OPERATIONS D ¹ ¹ A NUCLEAR REACTOR TO AVOID DAMAGE ¹ INTERACTION TABLET-LINER

The present invention relates to a method for determining operating limit values of a nuclear reactor to prevent damage due to the pastille-sheath interaction.

The invention is particularly, but not exclusively, applicable to pressurized water nuclear reactors (PWRs).

FIG. 1 diagrammatically illustrates such a pressurized water nuclear reactor 1 which conventionally comprises:

- a heart 2,

steam generators 3 (one per loop), a single steam generator being shown,

a turbine 4 coupled to a generator 5 of electrical energy, and

- a condenser 6.

The reactor 1 also comprises a primary circuit 8 equipped with pumps 9 (a pump per loop), a single pump being shown, in which water circulates under pressure, according to the path shown by the arrows. This water goes back particularly to the heart 2 to be warmed by ensuring the refrigeration of the heart 2. The water also provides a moderating function, that is to say, slowing the neutrons produced by the nuclear fuel. The primary circuit 8 further comprises a pressurizer 10 for regulating the pressure of the water circulating in the primary circuit 8.

The water of the primary circuit 8 also feeds the steam generators 3 where it is cooled by providing the vaporization of water circulating in a secondary circuit 12. The steam produced by the generators 3 is channeled by the secondary circuit 12 to the turbine 4 and then to the condenser 6 where the steam is condensed by indirect heat exchange with cooling water circulating in the condenser 6. The secondary circuit 12 comprises downstream of the condenser 6 a pump 13 and a heater 14.

The core 2 comprises fuel assemblies 16 which are loaded into a tank 18. A single assembly 16 is shown in FIG. 1, but the core comprises a plurality of assemblies 16. The fuel assemblies 16 comprise formed nuclear fuel rods, conventionally, an alloy sheath, based on zirconium, containing a stack of nuclear fuel pellets based on uranium oxide or a mixture of uranium oxide and plutonium oxide. The reactor 2 comprises control clusters 20 which are arranged in the tank 18 above certain assemblies 16. A single cluster 20 is shown in FIG. 1, but the core 2 comprises several tens of control clusters 20, for example sixty.

The control clusters 20 can be moved by mechanisms 22 to be inserted into the fuel assemblies 16 they overhang.

Conventionally, each control cluster 20 comprises control rods of neutron absorbing material.

Thus, the vertical displacement of each cluster 20 makes it possible to regulate the reactivity of reactor 1 and allows variations in overall power.

P supplied by the core 2 from zero power to the nominal power PN, depending on the depression of the control clusters 20 in the fuel assemblies 16.

It can be useful, in fact, especially in countries like France where 80% of the electricity is produced by nuclear reactors, that the overall power supplied by the reactors varies in order to adapt to the needs of the electricity network that they supply; we then talk about network monitoring. In particular, it is desirable to be able to operate the reduced power reactors for a long period where the network demand is low, before returning if necessary to the nominal power PN.

When the nuclear reactor is operating at its nominal power PN, the nuclear fuel pellets is, according to the term used in the art, conditioned.

The packaging is essentially characterized by the closing of a radial clearance between the pellets and the sheath, due to the creep of the sheath and the swelling of the pellets. Thus, more specifically, the following steps will be distinguished when packaging each pellet:

- Under the effect of the pressure difference between the outside and the inside of the fuel rod, the sheath is deformed radially gradually by creep towards the inside of the rod. Moreover, the fission products, which are mainly retained in the pellet, induce a swelling thereof. During this phase, the stress on the sheath from the point of view of the stresses is solely due to the differential pressure existing between the outside and the inside of the pencil. The stresses in the sheath are compressive stresses; the contact between the pellet and the cladding begins at the end of a time interval which essentially depends on the local conditions of irradiation (power, neutron flux, temperature, etc.) and on the material of the cladding. In reality, the contact is established gradually over a period that begins with a soft contact followed by the establishment of a strong contact. The pressure of contact of the oxide of the pellet on the inner face of the sheath leads to an inversion of stresses in the sheath which become positive and tend to urge the sheath in tension;

- The swelling of the pellet continues and then imposes its deformation to the sheath to the outside. In established steady state, this expansion is slow enough so that the relaxation of the material of the sheath allows a balance of efforts in the sheath. Under these conditions, the level of tensile stresses is moderate (a few tens of MPa) and does not pose any risk with respect to the integrity of the sheath. Thus, in steady state, the risk of breakage of the sheath by pellet-clad interaction (IPG) is low because of the thermomechanical equilibrium present in the sheath at relatively low stress levels. On the other hand, the risk of breakage of the sheath occurs when the power supplied by the fuel rod varies greatly. Indeed, an increase in power causes an increase in the temperature in the pencil. Given the difference in mechanical characteristics (coefficient of thermal expansion, Young's modulus) and the temperature difference between the uranium oxide pellet and the zirconium alloy sheath, the pellet will expand more than the sheath and impose its deformation on the latter.

In addition, the presence of a gap between the sheath and the pellet of corrosive fission products, such as iodine, creates the conditions of stress corrosion. Thus, the deformation imposed by the pellet on the sheath during a transient, or variation, power can cause damage or rupture of the sheath, conventionally named by sheath failure IPG.

However, such a sheath failure is not admissible for reasons of safety since it could cause the release of fission products in the primary circuit of the reactor. Power transients, ie power variations, can occur during normal operation of the reactor, that is to say in the so-called category 1 situations. power may be required especially to adapt to the electrical energy needs of the network. More severe power transients can also occur in so-called category 2 accident situations, following, for example, an excessive increase in charge, an uncontrolled removal of cluster group (s), a dilution of boric acid or a fall. Clusters undetected. These different situations are described in particular in application FR-2,846,139.

The risk of sheath failure by IPG only occurs when there is a significant increase in the local linear power dissipated by the fuel pellets.

During an accidental Category 2 spike, the power achieved locally in the fuel can increase rapidly and be two to three times higher than the nominal power, this rapid increase in power, due to the increase in the power level and / or deformation of the power distribution, causes significant expansion of the pellets, thereby increasing the internal surface stresses of the sheath.

The risk of sheath failure by IPG also occurs during a rapid recovery to 100% of the nominal power (PN) following a prolonged operation at intermediate power (FPPI). Indeed, during an IPPF, the fuel tends to decondition. The deconditioning of the fuel is characterized by the reopening of the radial clearance between the pellets and the sheath as well as the resumption of the creep of the sheath. The rapid return to 100% of the nominal power causes a rapid increase in the internal stresses of the sheath to a value greater than the internal stresses in nominal operation at 100% before the FPPI, risking to break the fuel sheath.

In order to guarantee the integrity of the rods with respect to the rupture of the sheaths by IPG, the pressurized water nuclear reactors comprise means of automatic shutdown of the reactor when the maximum linear power of the fuel rods exceeds a limit threshold of power. , also known as the automatic reactor shutdown threshold (AAR).

In the current PWRs, this power limit threshold is dimensioned according to the local linear power leading to sheath failure during the largest increases in linear power, ie during accidental category spikes. 2 and / or fast power up following an FPPI.

However, during transients of class 1 (normal operating transients), the risk of sheath failure occurs for greater local linear power than in transients of category 2, the kinetics of evolution of the transient power of transients. categories 1 being slow enough to allow relaxation of the stresses by creep.

Thus, the current design of the power limit threshold does not make it possible to obtain greater flexibility in the operation of pressurized water nuclear reactors, in particular in network monitoring, the operating margin being too much reduced by the low value of the threshold of power. automatic shutdown of the reactor.

In this context, the invention aims to solve the problems mentioned by allowing a more flexible operation of the nuclear reactor. To this end, the invention proposes a method for protecting against fuel clad breaks by pellet-clad interaction of a pressurized-water nuclear reactor, said method comprising the steps of: determining the maximum linear power of the fuel rods and the derivative of said linear power as a function of time; determining a linear power threshold as a function of said derivative of said linear power; comparing said maximum linear power with said linear power threshold; triggering an alarm and / or a corrective action to stop the power increase if said maximum linear power is greater than or equal to said determined linear power threshold. The term linear power, the thermal power produced per unit of active length of the fuel assembly.

Thanks to the invention, it is possible to adapt the value of the threshold limit of linear power as a function of the evolution kinetics of the maximum linear power of the fuel rods. The maximum linear power in the fuel rods is determined from measurements of operating parameters provided by a core instrumentation formed by sensors, conventional calculation software, data storage means.

Thus, for a Category 2 accidental power transient, the linear power limit threshold will preferably be set at a low limit, substantially around 450 W / cm in order to protect against the significant risk of sheath failure by IPG, whereas for a transient of category 1 with a slower increase in the linear power, the threshold limit of linear power can be set at a higher value, substantially up to 590 W / cm, the risk of sheath failure by IPG being lower. The high limit threshold equivalent to a maximum linear power substantially equal to 590 W / cm makes it possible in particular to protect the reactor against the melting of the fuel.

For this purpose, a correlation curve of the linear power threshold as a function of the derivative of the maximum linear power is stored in the storage means of the instrumentation of the heart.

According to another characteristic, the method is such that the corrective action is a shutdown of said nuclear reactor.

According to another characteristic, the method is such that it comprises the steps of: determining an alarm limit threshold of the maximum linear power; comparing said maximum linear power of the fuel rods with said limit alarm threshold of the maximum linear power; - trigger an alarm if said maximum linear power is greater than or equal to said alarm limit threshold of the maximum linear power.

Advantageously, the method comprises a step of decreasing the power of the reactor when said alarm of said alarm limit threshold of the maximum linear power is triggered.

Advantageously, said alarm limit threshold is determined by lowering said linear power threshold and intervenes before said linear power threshold (PL).

Advantageously, said alarm limit threshold of the maximum linear power corresponds to 94% of the linear power threshold.

According to another characteristic, the method is such as the step of determining said linear power threshold (PL) as a function of said

derived from said linear power () is realized by means of a dt linear function. The subject of the invention is also a protection device based on measurements of the internal and / or external instrumentation of the core implementing the method according to the invention, characterized in that it comprises: means for determining the linear power maximum (P * max) fuel rods (24) and the derivative of said power

..,. . ,. . dPC max. linear as a function of time (); and means for determining a linear power threshold (PL) in

function of said derivative of said linear power (); dt - means for comparing said maximum linear power {Pc max) with said linear power threshold (PL); means for triggering an alarm and / or a corrective action for stopping the power increase if said maximum linear power {Pc max) is greater than or equal to said determined linear power threshold (PL).

According to another characteristic, said device comprises control cluster control means for reducing the power of the core and / or stopping the reactor.

According to another characteristic, said device comprises storage means for storing the function making it possible to determine said linear power threshold (PL) as a function of said derivative of said power

..,. , dPC max. linear (). dt

Other features and advantages of the invention will emerge more clearly from the description which is given below, by way of indication and in no way limitative, with reference to the appended figures, among which:

- Figure 1 schematically illustrates a nuclear reactor with pressurized water;

FIG. 2 represents a fuel assembly of a pressurized water nuclear reactor; FIG. 3 represents a nuclear fuel rod of the fuel assembly illustrated in FIG. 2; FIG. 4 represents a block diagram of the process for determining operating limit values for a nuclear reactor according to the invention;

FIG. 5 is a graphical representation illustrating an exemplary correlation law of the process for determining operating limit values for a nuclear reactor according to the invention.

In all the figures, the common elements bear the same reference numbers.

Figure 1 has already been described above with reference to the general presentation of the invention.

FIG. 2 represents a fuel assembly 23 of a pressurized water nuclear reactor.

The fuel assembly 23 conventionally comprises an array of fuel rods 24 and a skeleton 26 for supporting the rods 24.

The skeleton 26 comprises, in a conventional manner:

a lower nozzle 28 and an upper nozzle 30 disposed at the longitudinal ends of the assembly 23,

- Guide tubes 21, connecting the two ends 26 and 28, for receiving control rods of the clusters 20,

- spacer grids 32 for holding the rods 24.

As illustrated in FIG. 3, each fuel rod 24 comprises a sheath 33 in the form of a tube closed at its lower end by a lower plug 34 and at its upper end by an upper plug 35. The pencil 24 contains a series of pellets 36 stacked in the sheath 33 and bearing against the lower plug 34. A holding spring 38 is disposed in the upper portion of the sheath 33 to bear on the top cap 35 and the top pad 36. Generally, the pellets 36 are made of uranium oxide and the sheath 33 is made of zirconium alloy.

In FIG. 3, which corresponds to a fuel rod 24 before irradiation, there is a radial clearance J between the pellets 36 and the sheath 33, for example between 100 and 300 μm. This is illustrated more particularly on the enlarged circled portion of FIG.

Each pellet 36 has a substantially cylindrical shape with chamfers (not shown) between end faces and its side face. A recess 37 spherical cap is formed in each end face, substantially in the center thereof.

In order to guarantee a good thermal exchange in the rods 24 before the pellets 36 come into contact with the sheath 33 during the conditioning of the pellets 36, the rod 24 is, in addition, filled with a thermally conductive gas, such as 'helium. The gas pressure also contributes to defer, in time, the compressive creep of the sheath 33.

When the reactor will operate, for example at its nominal overall power PN, the nuclear fuel pellets 36 will be conditioned.

The packaging is characterized essentially by the closing of the clearance J between the pellets 36 and the sheath 33, due to the creep of the sheath 33 and the swelling of the pellets 36.

Thus, more specifically, the following steps are distinguished for each pellet 36 during conditioning:

under the effect of the pressure difference between the outside (water of the primary circuit) and the inside of the pencil 24, the sheath 33 is progressively deformed radially by creep towards the inside of the pencil 24. Moreover, the products fission, which are mainly retained in the pellet 36, induce swelling thereof. During this phase, the stress on the sheath 33 from the point of view of the constraints is only due to the differential pressure existing between the outside and the inside of the pencil 24. The stresses in the sheath 33 are compressive stresses;

the contact between the pellet 36 and the sheath 33 starts at the end of a time interval which depends essentially on the local conditions of irradiation (power, neutron flux, temperature, etc.) and on the material of the sheath. In reality, the contact is gradually established over a period which begins with a soft contact followed by the establishment of a strong contact. The contact pressure of the oxide of the pellet 36 on the inner face of the sheath 33 leads to an inversion of the stresses in the sheath 33 which become positive and tend to urge the sheath 33 in tension;

- The swelling of the pellet 36 continues and then imposes its deformation to the sheath 33 to the outside. In established steady state, this expansion is slow enough for the relaxation of the sheath material 33 to allow an equilibrium of the forces in the sheath 33. Under these conditions, the level of the tensile stresses is moderate (a few tens of MPa) and presents no risk vis-à-vis the integrity of the sheath 33.

During a power transient category 2, the temperature of the pellets 36 can exceed widely 1200 ° C. This temperature corresponds to the activation domain of the thermal creep of the uranium-based fuel. Under the effect of the high temperature, the material of the pellets 36 tends to flow towards the center of the pellets 36 placed in compression, and thus to fill the recesses 37. Consequently, the stresses on the lateral surfaces of the pellets 36 , which comparatively remain cold (temperature below 1000 ° C), will decrease and thus the thrust of the pellets 36 on the sheath will also be reduced.

During this category 2 power transient, the sheath 33 imposes a feedback on the lateral faces of the pellets 36. temperature field prevailing within the fuel, the side faces of the pellets 36 remain relatively cold and therefore remain fragile and likely to be microcracked under the effect of the feedback imposed by the sheath 33. Thus, during a transient category 2 Counter-reactions 36-sheath pellets 33 can be established, both opposing the instantaneous effects of thermal differential shifts leading to a risk of rupture by IPG.

As a reminder, Category 2 transients, which are limiting to the risk of IPG breaking, are due in particular to one of the following events:

- excessive load increase,

- uncontrolled withdrawal of group (s) from clusters 20,

- cluster drop (s) 20, - dilution of boric acid.

It is known that accidental category 2 transients induce stronger and faster power variations in the core 2 than class 1 transients.

In order to guarantee the integrity of the rods 24 vis-à-vis the inter-chip interface 36-sheath 33, the nuclear reactor 1 comprises protection means adjusted so as to control the operation of the reactor 1 so that, in particular when a transient category 2, the maximum linear power Pe max in the rods 24 remains below a limit threshold PL.

The effective linear power Pc in the rods 24 is estimated, from the measurement of operating parameters provided by sensors, by conventional calculation software stored in a memory of the protection means.

In nuclear reactors with pressurized water, the means of protection are notably realized by an Integrated Protection System Digital, also called SPIN, or by an equivalent system based on measurements of the internal and / or external instrumentation of the reactor core.

For example, SPIN is formed by a set of electronic and electrical equipment that, based on measurements from instrumentation chains and comparing them to the limit thresholds, generates alarm signals, automatic power reduction or automatic shutdown of the reactor.

The method according to the invention makes it possible in particular to improve these control means and in particular devices of the SPIN or other equivalent type, based on the measurement of the internal and / or external instrumentation of the heart.

The method of determining operating limit values of a nuclear reactor to avoid damage due to the pastille-gaine interaction is illustrated by the block diagram of FIG. 4. The first step illustrated by block 40 consists of to determine the maximum linear power Pc max of the fuel rods 24 by means of the measurement of operating parameters provided by the sensors and by the calculation software stored in a memory of the protection means. In a second step, illustrated by block 41, the derivative is calculated

of the maximum linear power with respect to time, in order to evaluate the kinetics of evolution of the linear power in the rods 24. This derivative of the linear power makes it possible to provide the means of protection with information on the type of transient, category 1 or 2, sustained by the nuclear reactor in addition to the maximum value of the linear power of the fuel rods. This second step is performed simultaneously with the determination of the maximum linear power Pc max. In a third step, illustrated by block 43, a linear power limit threshold PL is determined as a function of the value of the derivative of the linear power determined previously. For this, a variation law, illustrated by block 42, is stored in the memory of the protection means. An example of a law is illustrated particularly by the graph of Figure 5.

According to a first embodiment of the invention, the correlation law between the derivative of the linear power as a function of time and the power limit threshold PL is represented by a linear function varying between two extreme thresholds, a high threshold and a low threshold.

Thus, as a function of the kinetics of evolution of the linear power, the variation law makes it possible to adapt the limit threshold of power PL as a function of the more or less rapid increase in the linear power of the rods 24; that is, depending on the severity of the transients. For an accidental category 2 transient involving a rapid increase in the linear power Pc, ie for a derivative of the linear power as a function of time, fuel rods leading to a high risk of rupture by IPG, the threshold limit of linear power is low, preferably of the order of 450 W / cm. For a normal operating transient, of category 1 with a linear power increase Pc slow, that is to say for a derivative of the linear power as a function of time, the risk of sheath failure by IPG at the threshold 450 W / cm is non-existent, the threshold power limit for this type of transient is greater, preferably of the order of 590W / cm. This threshold substantially equal to 590 W / cm makes it possible in particular to protect the reactor against the melting of the fuel.

The next step, illustrated by block 44, consists in comparing the value of the maximum linear power Pc max determined during the first step with the linear power limit threshold PL depending on the kinetics. Evolution of the linear power of fuel rods 24.

Block 45 illustrates the case where the maximum linear power Pc max is below the power limit threshold PL, in this case the protection means do not initiate corrective action to reduce the overall power P of the reactor.

On the other hand, if the maximum linear power Pc max is greater than the power limit threshold PL, then the protection means trigger either an alarm or a corrective action (typically a control cluster insertion) making it possible to reduce the overall power P of reactor, or even the automatic shutdown of the reactor.

Indeed, the power limit threshold PL is preferably determined in order to correspond to an automatic reactor shutdown (AAR).

In order to guard against an automatic shutdown of the reactor, a second threshold, called the alarm limit threshold of the maximum linear power SA, is dimensioned in order to prevent the approach of the power limit threshold PL corresponding to an arresting action. emergency reactor.

The alarm limit threshold of the maximum linear power SA is below the power limit threshold PL and advantageously corresponds to the power limit threshold PL minus a certain percentage, for example 6%.

For this purpose, the method according to the invention also comprises the steps of:

determining the alarm limit threshold of the maximum linear power SA from the linear power limit threshold PL;

- compare the maximum linear power of the fuel rods with the alarm limit threshold of the maximum linear power SA.

Thus, the subject of the invention is a method for determining operating limit values for a nuclear reactor in order to prevent damage. sheaths by pastille-sheath interaction.

The method according to the invention is implemented by a device for protecting the reactor core based on measurements of the internal and / or external instrumentation of the reactor core. The method according to the invention makes it possible to relax the power limit threshold as well as the resulting alarm threshold and makes it possible to increase the operational flexibility of the reactor while guaranteeing an effective protection against cladding failures by IPG. during accidental transients category 2. Of course, the invention is not limited to the embodiment just described.

Claims

1. A method of protection against fuel clad fractures by pastille-sheath interaction of a pressurized water nuclear reactor, characterized in that it comprises the steps of: determining the maximum linear power (P * max) of the rods fuels (24) and the derivative of said linear power in

, .. . . dP max. function of time (); dt - determine a linear power threshold (PL) as a function of

. ... ,. ,. . ... ..,. , <iP * max. said derivative of said linear power (); comparing said maximum linear power (P * max) with said linear power threshold (PL); trigger an alarm and / or corrective action to stop the power increase if said maximum linear power

(P * max) is greater than or equal to said determined linear power threshold (PL).

2. Method according to claim 1 characterized in that said corrective action is a judgment of said nuclear reactor.

3. Method according to claim 2 characterized in that said method comprises the steps of: determining an alarm limit threshold of the maximum linear power (SA); comparing said maximum linear power (P * max) of the fuel rods (24) with said alarm limit threshold of the maximum linear power (SA); triggering an alarm if said maximum linear power (P * max) is greater than or equal to said alarm limit threshold of the maximum linear power (SA).

4. Method according to claim 3 characterized in that it comprises a step of decreasing the power of the reactor when said alarm of said alarm limit threshold of the maximum linear power (SA) is triggered.

5. Method according to one of claims 3 to 4 characterized in that said alarm limit threshold of the maximum linear power (SA) is determined by lowering said linear power threshold (PL) and intervenes before said threshold of linear power. (PL).

6. The method as claimed in claim 5, characterized in that said alarm limit threshold of the maximum linear power (SA) corresponds to 94% of said linear power threshold (PL).

7. Method according to one of claims 1 to 6 characterized in that the step of determining said linear power threshold (PL) in

function of said derivative of said linear power () is realized by means of a linear function.

8. Protection device based on measurements of the internal and / or external instrumentation of the core implementing the method according to one of claims 1 to 7 characterized in that it comprises: means for determining the maximum linear power (P * max) fuel rods (24) and the derivative of said

linear power as a function of time (); dt means for determining a linear power threshold (PL) as a function of said derivative of said linear power dPt max; and means for comparing said maximum linear power (P * max) with said linear power threshold (PL); means for triggering an alarm and / or a corrective action for stopping the power increase if said maximum linear power (P * max) is greater than or equal to said determined linear power threshold (PL).

9. Protection device according to claim 8 characterized in that it comprises control cluster control means (s) for reducing the power of the heart and / or stop the reactor.

10. Protection device according to one of claims 8 to 9 characterized in that it comprises storage means for storing the function for determining said linear power threshold (PL) in

function of said derivative of said linear power (). dt