Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C7/00—Control of nuclear reaction
  • G21C7/06—Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section
  • G21C7/22—Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section by displacement of a fluid or fluent neutron-absorbing material, e.g. by adding neutron-absorbing material to the coolant
• • G—PHYSICS
  • G05—CONTROLLING; REGULATING
  • G05B—CONTROL OR REGULATING SYSTEMS IN GENERAL; FUNCTIONAL ELEMENTS OF SUCH SYSTEMS; MONITORING OR TESTING ARRANGEMENTS FOR SUCH SYSTEMS OR ELEMENTS
  • G05B13/00—Adaptive control systems, i.e. systems automatically adjusting themselves to have a performance which is optimum according to some preassigned criterion
  • G05B13/02—Adaptive control systems, i.e. systems automatically adjusting themselves to have a performance which is optimum according to some preassigned criterion electric
  • G05B13/04—Adaptive control systems, i.e. systems automatically adjusting themselves to have a performance which is optimum according to some preassigned criterion electric involving the use of models or simulators
  • G05B13/041—Adaptive control systems, i.e. systems automatically adjusting themselves to have a performance which is optimum according to some preassigned criterion electric involving the use of models or simulators in which a variable is automatically adjusted to optimise the performance
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C7/00—Control of nuclear reaction
  • G21C7/36—Control circuits
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21D—NUCLEAR POWER PLANT
  • G21D3/00—Control of nuclear power plant
  • G21D3/001—Computer implemented control
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21D—NUCLEAR POWER PLANT
  • G21D3/00—Control of nuclear power plant
  • G21D3/08—Regulation of any parameters in the plant
  • G21D3/10—Regulation of any parameters in the plant by a combination of a variable derived from neutron flux with other controlling variables, e.g. derived from temperature, cooling flow, pressure
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21D—NUCLEAR POWER PLANT
  • G21D3/00—Control of nuclear power plant
  • G21D3/08—Regulation of any parameters in the plant
  • G21D3/12—Regulation of any parameters in the plant by adjustment of the reactor in response only to changes in engine demand
  • G21D3/16—Varying reactivity
• • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
  • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
  • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
  • Y02E30/00—Energy generation of nuclear origin
  • Y02E30/30—Nuclear fission reactors

Definitions

• TITLE Method and assembly for controlling a nuclear reactor, nuclear reactor equipped with such an assembly
• the invention generally relates to the control of nuclear reactors, in particular during power transients.
• a market price for electricity below zero should push nuclear reactor operators to rapidly reduce the electrical power produced by the reactor. On the other hand, a rise to high power is desired when electricity prices become positive again.
• mode A Most of the PWR (Pressurized Water Reactor or REP) type nuclear reactors in the world are operated in a mode generally referred to as mode A. These reactors are designed for base operation, i.e. ie at a high and substantially constant electric power. Power variations are tricky to achieve for reactors operating in mode A, and require precise actions on the part of the operators, so as to avoid the triggering of alarms or protections.
• REP Pressure Water Reactor
• a third possibility is to switch from mode A to an RMOSC-type operating method. This process is protected by the patent application filed under number PCT/EP2019/052543, in the name of the applicant.
• This method implements a core control means with two regulators in cascade, a supervisor using a predictive control algorithm, and a multi-objective regulator.
• the invention relates, according to a first aspect, to a method for controlling a nuclear reactor, the nuclear reactor having a core comprising a plurality of nuclear fuel assemblies, a primary core cooling circuit in which a primary heat transfer fluid containing a neutron poison, a unit making it possible to inject neutron poison into the primary heat transfer fluid, and a unit provided for injecting water into the primary circuit, the method comprising the following steps:
• S20/ acquisition of current values of a plurality of operating parameters of the nuclear reactor comprising at least one parameter characterizing a supplied core power by the core of the reactor and a parameter characterizing the distribution of the neutron flux in the core;
• steps S20/ and S30 being repeated with a period of less than 60 minutes.
• This process makes it possible to quickly determine an optimum sequence for injecting neutron poison and/or water into the primary coolant, allowing operators to carry out the power transient with confidence.
• a key point is that the injection sequence is determined repeatedly, with a period of less than 60 minutes, for example of the order of ten minutes.
• the injection sequence is therefore periodically updated based on the acquired values of the operating parameters of the nuclear reactor, for example every ten minutes.
• control method may also have one or more of the characteristics below, considered individually or in all technically possible combinations:
• the at least one quantity characteristic of the state of the core calculated in step S30/ comprises said parameter characterizing the distribution of the neutron flux in the core
• the cost function characterizes an evolution of a difference between said parameter characterizing the distribution of the neutron flux in the core and a reference value over said determined time interval
• the convergence criterion includes reaching an extremum of the cost function
• the convergence criterion includes satisfying at least one constraint chosen from the following list:
• the neutron poison and/or water injection sequence in the primary liquid is generated by considering the results obtained in the previous iteration, by a gradient descent algorithm,
• step S30/ comprises a sub-step S35/ of determining an optimum slope for a change in power as a function of time during the power variation from the first power to the second power, the sub-step S35/ comprising the following operations:
• the predictive model of the reactor core is non-linear.
• the predictive model of the core of the reactor comprises several sub-models, each sub-model modeling a level of the core of the nuclear reactor and comprising at least one equation describing a kinetics of a density of neutrons at said level and an equation describing a temperature of the primary heat transfer fluid at said level, the model further comprising equations describing neutron exchanges between the levels and equations characterizing a reactivity at each level.
• the neutron poison and/or water injection sequence in the primary liquid comprises a plurality of injection operations, each operation being characterized by a quantity and an operation duration, the number of operations in the injection sequence being between 2 and 12, the duration of operation being between 2 minutes and 60 minutes.
• the period T is less than or substantially equal to the duration of an operation of the injection sequence.
• the determined time interval has a total duration between 10 minutes and a duration of the power program.
• the invention relates to a control unit for a nuclear reactor, the nuclear reactor having a core comprising a plurality of nuclear fuel assemblies, a primary core cooling circuit in which circulates a primary heat transfer fluid containing a neutron poison, a unit for injecting neutron poison into the primary heat transfer fluid, and a unit provided for injecting water into the primary circuit, the neutron poison and water injection units being controlled by an operator
• the control assembly comprising: a/a user interface, configured so that a user enters a power program to be supplied by the nuclear reactor, this program comprising at least one power variation from a first power to a second power; b/ a unit for acquiring current values of a plurality of operating parameters of the nuclear reactor, comprising at least one parameter characterizing a power supplied by the reactor core and a parameter characterizing the distribution of the neutron flux in the core; c/ a calculation unit comprising:
• an optimization algorithm programmed to generate a neutron poison and/or water injection sequence in the primary liquid covering a determined time interval
• a predictive model of the core of the reactor programmed to calculate an evolution of at least one quantity characteristic of the state of the core of the nuclear reactor during the said time interval determined by using the power program acquired, the current values of the operating parameters acquired and the injection sequence considered;
• a cost module configured to calculate a cost function, using the evolution calculated by the predictive model; the optimization algorithm being programmed to, iteratively, generate an injection sequence, have the evolution of the at least one corresponding characteristic quantity calculated by the predictive model of the heart, have the corresponding cost function evaluated by the cost modulus, until a convergence criterion of the cost function is satisfied; the assembly is also configured to display the optimum injection sequence, i.e. the one for which the convergence criterion of the cost function has been reached, on the user interface, so that the sequence of optimum injection is implemented by the operator.
• control assembly is such that the calculation unit comprises a slope module programmed to determine an optimum slope for a change in reactor power as a function of time during the power variation from the first power to the second power, said module being programmed for:
• the invention relates to a nuclear reactor comprising a core comprising a plurality of nuclear fuel assemblies, a primary core cooling circuit in which circulates a primary heat transfer fluid containing a neutron poison, a unit making it possible to inject neutron poison into the primary heat transfer fluid, a unit provided to inject water into the circuit primary, and a control assembly having the above characteristics, the neutron poison and water injection units being controlled by an operator.
• FIG. 1 is a schematic representation of a reactor nuclear connected to the electrical distribution network, equipped with the steering assembly of the invention
• Figure 2 is a step diagram illustrating the method of the invention
• FIG. 3 is a schematic representation of the predictive model of the reactor core used in the method of FIG. 2
• FIG. 4 is a schematic representation illustrating the injection sequence and the evolution of the parameters calculated according to the method of the invention
• FIG. 5 is a graphic representation of the evolution of the axial power imbalance (axial offset AO in English) obtained when the method of the invention is implemented, for a power program comprising successively a drop in power followed by a rise in power, at the start of the cycle;
• FIG. 6 is a graphical representation similar to that of FIG. 5, showing the evolution over time of the difference between the average temperature of the primary coolant in the core Tmoy and the reference temperature Tref, for the same scenario as Figure 5;
• FIGS. 7 and 8 are graphical representations similar to those of FIGS. 5 and 6, for the same scenario, at the end of the cycle; and
• FIG. 9 is a simplified diagrammatic representation of the various modules constituting the control assembly equipping the reactor of FIG. 1.
• the reactor 1 shown in Figure 1 comprises, in a conventional manner, a core 3, itself comprising a plurality of nuclear fuel assemblies 5.
• the nuclear reactor 1 also comprises a primary circuit 7, provided for cooling the core 3, in which circulates a primary heat transfer fluid containing a neutron poison.
• this primary circuit comprises several loops, each loop having a steam generator 9 and a primary pump 11.
• the nuclear reactor also comprises a secondary circuit 13, in which circulates a secondary heat transfer fluid.
• the secondary heat transfer fluid is vaporized in the steam generator 9, under the effect of the heat released by the primary heat transfer fluid.
• the secondary circuit 13 comprises at least one turbine 15, a condenser 17, a supply tank 19, and secondary pumps 21, 23.
• the secondary heat transfer fluid in the form of vapor circulates from the steam generator 9 to the turbine 15 then is condensed in the condenser 17. It is then returned in liquid form to the steam generator 9.
• a valve 25 is interposed on the steam line connecting the steam generator 9 to the turbine 15, and makes it possible to adjust the flow rate of steam supplying the latter.
• the turbine 15 mechanically drives an alternator 27.
• the electricity generated by the alternator 27 feeds an electrical distribution network 29.
• the nuclear reactor 1 also includes a unit 31 for injecting neutron poison into the primary heat transfer fluid.
• Neutron poison is typically boron.
• a pump 35 and a valve 37 are interposed on line 33.
• Unit 31 selectively increases the concentration of neutron poison in the primary liquid.
• the nuclear reactor also includes a unit 39 designed to inject water into the primary circuit.
• Unit 39 comprises a tank 41 connected by a line 43 to the primary circuit 7.
• a pump 45 and a valve 47 are inserted in line 43.
• the water is typically pure demineralised water.
• Unit 39 is provided for injecting water into primary circuit 7, which has the effect of reducing the concentration of neutron poison in the primary heat transfer fluid.
• the nuclear reactor 1 further comprises groups 49 of control rods and a mechanism 51 capable of selectively inserting or extracting the groups of control rods 49 in the core 3 of the nuclear reactor.
• the control rods are made of a neutron absorbing material.
• Groups of control rods are 49 moved in such a way as to selectively alter the responsiveness inside Core 3.
• Nuclear reactor 1 also includes an instrumentation and control system 53.
• This instrumentation and control system 53 comprises instrumentation 55 making it possible to directly measure or determine a plurality of operating parameters of the nuclear reactor. These operating parameters include at least the following parameters:
• the instrumentation 55 comprises in particular neutron detectors, placed outside the core 3, and distributed over the entire height of the core. These detectors are known as ex-core chambers.
• the power supplied by the core is for example obtained by calculation, on the basis of the information provided by the detectors measuring the neutron fluxes outside the core.
• the power delivered by the core is determined using the measurement of the power delivered by the turbine.
• the parameter characterizing the distribution of the neutron flux in the core is for example the axial power distribution, or axial offset AO.
• the axial offset is calculated using the following formula:
• AO (0h-0b)/(0h+0b) where 0h is the neutron flux from the upper half of the core, and 0b is the neutron flux from the lower half of the core.
• the neutron fluxes from the upper and lower halves of the core are typically obtained by the neutron detectors placed outside the core 3. Alternatively, they are obtained by probes placed in the core 3, called in-core probes.
• the parameter characterizing the distribution of the neutron flux in the core is FIi-F ⁇ , or any other suitable parameter.
• the current flow rate of neutron poison and the current flow rate of water injected into the primary heat transfer fluid are also measured or determined.
• the instrumentation and control assembly 53 further comprises a control device 57 configured to regulate a certain number of operating parameters of the nuclear reactor.
• the control device 57 comprises at least one loop 59 for controlling the temperature of the primary heat transfer fluid.
• the loop 59 receives as input the average current temperature Tmoy of the primary heat transfer fluid in the core 3 of the reactor.
• This value corresponds for example to the average of the temperature measured at the reactor inlet of the temperature measured at the core outlet.
• the average temperature of the primary coolant Tmoy in the core is controlled by moving groups of control bars 49.
• Four groups of bars, called groups A, B, C, D, can be moved to control the temperature Tmean.
• the control device 57 also comprises a loop 61 for controlling the power supplied by the turbine 15.
• Loop 61 receives as input the value of the power supplied by turbine 15. Loop 61 also receives a turbine power setpoint and a slope setpoint for any variation in turbine power.
• the power and slope setpoints are typically set by the operator of the nuclear reactor.
• the loop 61 controls the valve 25 interposed on the steam line of the secondary circuit 13, according to the power and slope setpoints, and according to the current value of the turbine power.
• the operator directly controls the neutron poison and water injection units 31 and 39.
• It sets the quantity of neutron poison injected per unit time into the primary heat transfer fluid and the quantity of water injected per unit time into the primary heat transfer fluid. Typically, it sets the volume flow rate of the neutron poison solution injected, and the volume flow rate of water injected.
• the invention is particularly suited to the case where the nuclear reactor must follow a power program comprising at least one power variation, also called transient, from a first power to a second power.
• the power program in this case is provided to the operator of the nuclear power plant by the person responsible for managing the electricity transmission network 29.
• the average temperature of the primary coolant Tmoy in the core is controlled by moving groups of 49 control rods.
• Variations in the level of power supplied by the nuclear reactor are achieved by adjusting the concentration of the neutron poison in the primary coolant. This is done by injecting neutron poison or water into the primary circuit 7, using units 31 and 39.
• concentration of the neutron poison is increasingly slow as the exhaustion of the nuclear fuel progresses.
• the maximum load variation slope is 1.5% of nominal power per minute at the start of the cycle, and 0.10% of nominal power per minute at 90% of the cycle.
• these power adjustments are tricky to make because they require very precise control of the quantities of neutron poison injected into the primary circuit.
• the maximum slope possible in G-mode or T-mode is 5% of nominal power per minute up to 80% of the cycle.
• the invention aims to overcome these difficulties by adding to the nuclear reactor a control assembly 63 which will provide the operator with a sequence of injection of neutron poison and/or water into the primary liquid, specially adapted for the program of power to be supplied.
• the control unit 63 also provides a recommended slope for the or each power variation to be performed during the power program.
• the steering assembly 63 is provided to implement the method for steering the nuclear reactor which will now be described.
• the control method comprises, as illustrated in FIG. 2, a step S10 for acquiring a reactor power program to be supplied by the nuclear reactor.
• This program includes at least one reactor power variation from a first power to a second power.
• the reactor power typically corresponds to the mechanical power supplied by the turbine.
• the reactor power variation is typically of non-zero amplitude.
• control method is particularly suited to the case of a nuclear reactor following a power program having a power transient.
• the power program to be followed is in this case typically provided by the manager of the electricity distribution network 29, as described above.
• the control method is also suitable for reactors not operating in load monitoring but having to manage significant power transients.
• the variation in reactor power is typically several tens of percent of the nominal power of the reactor.
• the control method also applies to power variations of small amplitudes, for example when the reactor operates in remote control.
• the power variations are then a few percent of the nominal power of the reactor, for example less than 10%, or even less than 5%.
• the control process still applies to the case where the nuclear reactor is operating on base.
• the power variations are then zero, the first power being equal to the second power.
• the nuclear reactor operates at constant power, typically at 100% of its nominal power (PN).
• PN nominal power
• the exhaustion of the fuel in such a case causes a drop in the average temperature of the primary coolant, which causes a modification of the cost function.
• the control process proposes recommendations for the injection of neutron poison or water to restore an optimal situation.
• the reactor power program covers a twenty-four hour period. It comprises a single reactor power variation, or may alternatively comprise several.
• the reactor power program is a slot, first comprising a drop in power, followed a few hours later by a rise to the initial power level.
• the control method also includes a step S20 for acquiring the current values of a plurality of operating parameters of the nuclear reactor 1.
• the operating parameters include at least one parameter P characterizing a core power supplied by the core of the nuclear reactor 3, and a parameter R characterizing the distribution of the neutron flux in the core 3.
• step S20 one or more of the following operating parameters are also acquired in step S20:
• the parameter characterizing the core power P supplied by the core of the reactor is for example the thermal power supplied by the core.
• This parameter is reconstructed by system 53 using neutron flux measurements from neutron detectors located outside the core.
• this parameter is reconstituted from the measurement of the power supplied by the turbine, or the temperatures of the primary coolant at the inlet and at the outlet of the Tin and Tout core.
• the parameter characterizing the core power is the total neutron flux in the core, or the power supplied by the turbine, or any other suitable parameter.
• the parameter R characterizing the distribution of the neutron flux in the core is typically the axial offset AO. It is typically reconstituted, as described above, from the measurements provided by the probes measuring the neutron flux outside the core or inside the core. Alternatively, this parameter is the difference between the neutron flux in the upper part of the core and the neutron flux in the lower part of the core.
• the mean temperature Tmoy of the primary coolant is calculated using the measurement of the temperature of the primary coolant at the core outlet Tout and the measurement of the primary coolant fluid at the core inlet Tin.
• Tavg is calculated using the following equation:
• Taver (Tin + Tout)/2
• the quantity of neutron poison injected per unit time Qpn is determined by using a flow sensor installed on the conduit 33. As a variant, it is determined by using the speed of rotation of the rotor of the pump 35 or any other suitable quantity.
• the quantity of water injected per unit of time Qw is determined by using the measurement provided by a flow sensor installed on the conduit 43. Alternatively, it is reconstituted by using the speed of rotation of the rotor of the pump 55, or any other suitable size.
• the method further comprises a step S30 of determining the optimum neutron poison and/or water injection sequence, for a determined time interval, taking into account the reactor power program to be carried out.
• Step S30 includes several sub-steps, which are implemented iteratively.
• Step S30 includes a sub-step S31 for generating a neutron poison and/or water injection sequence in the primary liquid, covering a determined time interval.
• the neutron poison and/or water injection sequence comprises a plurality of injection operations, each operation being characterized by an amount of neutron poison or water injected, and an operation duration.
• the number of operations in the injection sequence is between two and twelve, preferably between three and eight, and is for example six.
• the duration of operation is between two and sixty minutes, preferably between five and twenty minutes, and is for example ten minutes.
• the determined time interval has a total duration comprised between ten minutes and the duration of the power program, preferably comprised between twenty minutes and three hours, more preferably comprised between thirty minutes and two hours and is worth for example one hour.
• the determined time interval covers a portion of the reactor power schedule.
• the quantity injected at each operation is expressed in volume, or in mass, or corresponds to a flow rate of neutron poison solution or water.
• the injection operations are successive operations, immediately following each other, covering together the entire time interval.
• Step S30 further comprises a sub-step S32 of calculating a change in at least one characteristic quantity of the state of the core of the nuclear reactor during said determined time interval.
• the at least one characteristic quantity of the calculated core state depends, among other things, on the chosen cost function, which will be described below.
• the at least one characteristic quantity comprises at least the parameter R characterizing the distribution of the neutron flux in the core, for example the axial offset AO.
• This calculation is performed using a predictive model of the core of the reactor, using the power program acquired, the current values of the operating parameters acquired, and the injection sequence generated in sub-step S31. More precisely, the portion of the power program covered by the determined time interval is used in sub-step S32.
• the predictive model of the reactor core is a nonlinear model.
• the model is linear. This model is then for example obtained by linearizing the nonlinear model described below.
• the predictive model of the heart includes several sub-models, each sub-model modeling a level of the heart 3.
• the core is divided, along the vertical direction, into several slices, each sub-model modeling one of the core slices.
• the predictive model of the heart comprises between two and twenty sub-models, preferably between two and ten sub-models, and comprises for example six sub-models.
• Each sub-model includes at least one equation describing a kinetics of a neutron density in said level, and an equation describing a temperature of the primary coolant at said level.
• the temperatures T2 to T7 at the exit of each level are deduced from the neutron flux in each level.
• - KT/H power/neutron flux conversion coefficient
• - Qp mass flow rate of primary coolant in the core.
• each level of the core is modeled using a one-group approximation of the point neutron kinetics with, in addition, a coefficient D to account for the neutron exchanges between the levels.
• Delayed neutrons and precursors which have a dynamic greater than the dynamic expected by the process or the operator, are not modeled. These neutrons have a typical dynamic of 10 seconds, for a calculation time step of the order of 60 seconds.
• the model also includes equations characterizing the reactivity at each level
• Equations are: p,o: initial reactivity at level i, determined when the initial state of the predictive model of the heart is adjusted on the basis of the current values of the acquired operating parameters (see below);
• thermal power supplied by level i of the core assumed to be proportional to the density of neutrons
• D denotes a variation with respect to the initial state of the predictive model.
• the effect due to the variation in the power supplied by the core makes it possible to characterize the effect of a variation in the temperature of the nuclear fuel, also called the Doppler effect.
• APbanki corresponds to the displacement of all the control groups within level i, expressed in steps. This parameter corresponds to the sum of the variations of the number of insertion steps in the level i of all the control groups.
• Kn DR, Kn x An, where Kn is a predetermined constant.
• the model incorporates the following general equation: P being the total power supplied by the core.
• the coefficients K T/H , K n K m0d , K dop , K b0r , K bank , K xenon , D, the evolution coefficients of iodine and xenon P , r xe and s ⁇ , were determined by numerical simulations. Some can also be determined by on-site measurements.
• the coefficients I*, h ⁇ , lc e are known values. I * alternatively is determined by calculation.
• the value of the APbank is determined by the model, according to the evolution of the mean temperature Tmoy of the primary coolant in the core. This module first determines the value of Taver according to the following equation:
• Tmean (T1 + T7)/2.
• the model determines an average reference temperature Tref, as a function of the reactor power supplied by the power program and the slope retained for the transients.
• the model determines the difference ATmoy between Tmoy and the mean reference temperature Tref.
• the model sets the value of Tmean at the limit of the dead band and determines that it is necessary to move the control groups . It calculates a value of APbank according to ATmoy, allowing to reach the criticality.
• the reference temperature values and the width of the temperature dead band as a function of reactor power are predetermined values, recorded in the model, or are provided by the instrumentation and control system 53.
• This model thus simulates the operation of the temperature loop 59.
• ACpn is obtained by integrating the quantities of neutron poison and/or water injected into the primary heat transfer fluid.
• Mt total mass of water in the primary circuit for example 260 tonnes for an N4 unit
• a delay is considered to assess the effect of the injection of demineralised water or neutron poison.
• the value of Qp i.e. the primary coolant flow rate in the core, is a predetermined value.
• step S30 Before the first iteration of step S30, that is to say immediately after the step of acquiring the current values of the operating parameters, the initial state of the predictive model of the heart is adjusted by using the current values of the parameters acquired operations.
• the values of the power released P by the heart and of Tmoy are used to set Ti at T7.
• the values of the power P released by the core and of the axial offset are used to determine the values of m to n 6 .
• the position value of the cgroups is used to directly set the starting value of Pbank.
• the equations are balanced in the model by adjusting the different reactivity terms, so that the time evolution of the neutron densities is zero.
• the predictive core model calculates for each level of the core the evolution over time, over the determined time interval, of the neutron concentration n,, of the temperature T, and of the xenon concentration Xe, .
• the model reconstructs the evolution of more global parameters such as the average temperature Tmoy, and the parameter R characterizing the neutron distribution in the core, for example the axial offset AO.
• the model also determines the evolution of the position of the Pbank groups and the concentration of neutron poison Cpn, as described above.
• the evolution of the power P emitted by the core follows the power program during the determined time interval, taking into account the slope retained for the power variations.
• Step S30 further includes a sub-step S33 for evaluating a cost function, using the evolution of the characteristic quantity(s) determined in sub-step S32.
• the cost function characterizes an evolution of a difference 5R between the parameter characterizing the distribution R of the neutron flux in the core, typically the axial offset, and a reference value Rref, over said determined time interval.
• Rref is a predetermined value, depending on the reactor power.
• the reference value Rref is entered manually by the operator. As a variant, it is recovered in the instrumentation and control system 53. Indeed, this system comprises a reference curve directly giving the reference value as a function of the current reactor power.
• Rref is taken constant over the determined time interval or variable according to the power program.
• the cost function is chosen to minimize the differences between the parameter R characterizing the distribution of the neutron flux in the core and the reference value.
• K can be assigned to the constant K depending on whether it is at the start of the cycle for the assemblies loaded in the reactor core, or at the end of the cycle. For example, if the parameter characterizing the neutron flux distribution in the core is the axial offset AO, the value retained for K is -2% AO/°C at the start of the cycle, -6% AO/°C at 80% of the cycle.
• the advantage of this second function is to allow the determination of the optimal injection sequence even when Tmoy remains inside its dead band. Indeed, in this case, the groups of control rods are not moved and the value of the parameter R characterizing the distribution of the neutron flux in the core does not change.
• the first cost function considered, in such a situation is constant regardless of the values of Q pn and Q w injected.
• the second cost function varies and makes it possible to discriminate between the different injection sequences considered.
• the second function therefore makes it possible to take into account the variations of the parameter R characterizing the distribution of the neutron flux in the core induced by the variations of Tmoy in its dead band.
• Step S30 then includes a sub-step S34 consisting in determining whether a convergence criterion of the cost function is satisfied. As illustrated in FIG. 2, if this convergence criterion is not satisfied, the sub-steps S31, S32 and S33 are repeated, with a new neutron poison and/or water injection sequence.
• the neutron poison and/or water injection sequence in the primary liquid is generated for the new iteration by considering the results obtained at the previous iteration.
• the injection sequence for the new iteration is generated using a gradient descent algorithm, typically with an optimization method known as the primal dual interior points method.
• step S40 is performed.
• the optimum injection sequence that is to say the one making it possible to reach the convergence criterion, is communicated to the operator.
• the injection sequence and the evolution of the characteristic quantities of the state of the heart during the determined time interval are displayed on a screen of the user interface, as represented in figure 3.
• the convergence criterion includes at least the fact that the cost function reaches an extremum.
• this extremum is a minimum.
• the difference between the parameter R characterizing the distribution of the neutron flux in the body and the reference value must be minimum over said determined time interval.
• the convergence criterion may provide that one or more of the following constraints are satisfied:
• the first constraint translates the fact that the parameter R characterizing the distribution of the neutron flux in the core, typically the axial offset AO, must not exceed limits around its reference value Rref during the determined time interval.
• This criterion can be expressed by the following equation:
• the limit is defined by the normal operating conditions of the nuclear reactor.
• the limit imposed on the quantity of neutron poison injected per unit time typically corresponds to the maximum volume flow likely to be delivered by unit 31.
• the limit imposed on the quantity of water injected per unit of time also corresponds to the maximum volume flow likely to be delivered by unit 39.
• steps S20 and S30 are repeated with a period T of less than sixty minutes, preferably less than twenty minutes, and for example equal to ten minutes.
• the method makes it possible to provide the operator with an optimum injection sequence, repeatedly with a short period.
• This injection sequence is reset to the current values of the reactor operating parameters.
• the period T is less than or equal to the duration of an operation of the injection sequence.
• the period T is a parameter adjustable by the operator of the nuclear reactor.
• the minimum value of T corresponds to the duration for executing steps S20 and
• the period T is substantially equal to the duration of an operation of the injection sequence.
• the operator launches the execution of steps S20 and S30 at t0. It implements between t0 and t0+T the setpoint determined in the previous iteration for the first operation of the optimum injection sequence.
• the operator receives the result of the new calculation, i.e. the new optimum injection sequence towards the end of the current injection operation, i.e. a little before t0+T . Between t0+T and t0+2T, it implements the recommendations received for the first operation of the new optimum injection sequence, and so on.
• the period T is less than the duration of an operation of the injection sequence.
• DI duration of an operation of the injection sequence.
• This variant is chosen when the calculation time for executing steps S20 and S30 is short.
• the calculation time can be less than 1 minute.
• the operator implements between t0 and t0+DI the setpoint determined at the previous iteration (between t0-DI and t0) for the first operation of the optimum injection sequence.
• the operator launches the execution of steps S20 and S30 at t0. These steps are repeated several times between t0 and t0+DI.
• the operator retains the result of the last calculation, i.e. the optimum injection sequence received towards the end of the injection operation in progress, a little before t0+DI.
• t0+DI and t0+2DI it implements the recommendations received for the first operation of the new optimum injection sequence, and so on.
• step S30 includes a sub-step S35 for determining an optimum slope for the evolution of the power as a function of time, during the power variation from the first power to the second power.
• This sub-step S35 is performed immediately before sub-step S31, as illustrated in Figure 4.
• the optimum slope that will be determined is the maximum feasible slope without the operator taking the risk of triggering an alarm or protection of the nuclear reactor.
• Sub-step S35 includes the following operations:
• slope values tested are the typical slope values for a nuclear reactor under the circumstances considered. These slope values are well known. They vary for example between 0.5 and 10%, typically between 0.1 and 5%, depending on the situation.
• the slope values are constant throughout the duration of the power variation.
• the slope values are variable during the transient. For example, different slope values can be chosen for different portions of the transient. Thus, a slope value can be chosen for the 80%PN-90%PN portion, another for the 90%PN-100%PN portion, etc.
• Operation S351 is performed with the predictive model of the heart already readjusted using the current values acquired in step S20.
• the calculation horizon here corresponds to the entire duration necessary to pass from the first power to the second power, taking into account the slope considered.
• the cost function used for operation S352 is the one described above.
• the slope value considered as optimum is that for which the cost function is minimum or maximum, depending on the cost function retained.
• the slope value considered as optimum is that minimizing the cost function.
• the optimum slope value is that which not only minimizes or maximizes the cost function, but also respects one or more operational constraints. These operational constraints are those listed above.
• the slope value retained for the power variation is communicated to the operator at step S40.
• the slope value retained corresponds to the optimum slope determined minus a safety factor.
• the safety factor is for example 30%.
• step S32 the evolution of the power of the core considered to carry out the simulations is determined by using the retained slope value, determined in sub-step S35.
• This slope is considered to be constant throughout the power transient. It is only modified if a new power program is proposed by the operator. Alternatively, as indicated above, the slope is variable during the transient.
• sub-step S35 is not implemented. In this case, it is the operator who sets the slope at which the power transient will be carried out from the first power to the second power. It is this value which will be considered for each iteration of sub-step S32.
• An optimum slope is determined or fixed for each variation in reactor power of the power program.
• FIGS. 5 to 8 illustrate the results obtained by implementing the piloting method of the invention.
• the core is loaded with fresh fuel assemblies, with a substantially zero burnout rate and a boron concentration of around 1200 ppm in the primary coolant.
• the simulation is performed for a core at 80% of its cycle, and a boron concentration of around 200 ppm.
• the power program considered is illustrated in figure 5 and in figure 7.
• the nuclear reactor operates in load following mode and follows a classic cycle in which the reactor power is first reduced from 100% of the nominal power to 50% nominal power, then increased to 100% of nominal power. A gap of eight hours separates the start of the load drop from the start of the load increase.
• the method of the invention indicates that a slope of 1.6% of the nominal power per minute is retained for the drop in load, and of 1% of the nominal power per minute for the ascent.
• Figure 5 shows the evolution of the axial offset in the reactor core when the optimum injection sequences are applied.
• Figure 6 shows the difference between Tmoy and Tref when the optimum injection sequences are applied.
• Figure 5 also shows the axial offset reference value (horizontal line in the center of the figure), as well as the limits of the axial offset dead band (horizontal line in dashed lines at the top and bottom of the figure 5).
• the limits in this non-limiting example, have been taken at 5%.
• FIG. 6 the limits of the temperature deadband T mean are shown in broken lines at the top and bottom of the figure.
• Figure 6 clearly shows that Tmoy remains in its dead band, and only comes out of it occasionally, during power transients, due to temperature regulation.
• Figures 7 and 8 are similar to Figures 5 and 6.
• the slopes adopted according to the method of the invention are 1.6% of the nominal power per minute for the load drop, and 0.3% of the nominal power per minute for the ascent.
• Figure 7 shows that the axial offset temporarily leaves its dead band, at the end of the load drop.
• this does not call into question the feasibility of this load transient, because this exit from the dead band is of short duration. Tmoy also comes out of its dead band from time to time, especially when the load drops.
• the steering assembly 63 which will now be detailed is particularly suitable for implementing the steering method described above.
• control unit 63 comprises a user interface 65, configured so that an operator informs the power program to be supplied by the nuclear reactor.
• the power program includes at least one power variation from a first power to a second power.
• the power program is as described above for the control method.
• the user interface 65 is of any suitable type. For example, it comprises a keyboard and a screen, connected to a computer.
• Assembly 63 also includes a unit 67 for acquiring current values of a plurality of operating parameters of the nuclear reactor.
• This unit 67 is a computer, or part of a computer.
• the acquisition unit 67 recovers the current values of the operating parameters in the instrumentation and control system 53.
• the plurality of operating parameters of the nuclear reactor comprises at least one parameter characterizing the power P supplied by the core of the reactor, and one parameter R characterizing the distribution of the neutron fluid in the core.
• the operating parameters acquired by unit 67 are those described above for the control process.
• the steering assembly 63 further includes a calculation unit 69.
• the calculation unit 69 comprises a predictive model 71 of the core of the reactor, a module 73 configured to calculate a cost function, and an optimization algorithm 75.
• the unit 67 is for example integrated into the calculation unit 69.
• the optimization algorithm 75 is programmed to generate a neutron poison and/or water injection sequence in the primary liquid covering a determined time interval.
• the sequence of injection of neutron poison and/or water into the primary liquid comprises a plurality of injection operations, each operation being characterized by an operation quantity and duration.
• the injection sequence is as described above for the piloting process.
• the predictive model of the core of the reactor 71 is programmed to calculate an evolution of at least one quantity characteristic of the state of the core of the nuclear reactor during the determined time interval, by using the acquired power program, the current values of the operating parameters acquired and the injection sequence considered.
• This predictive model 71 is the one described above.
• the predictive model 71 considers the portion of the power program corresponding to the time interval covered by the injection sequence.
• the at least one quantity characteristic of the state of the heart calculated by the predictive model 71 depends among other things on the chosen cost function. Typically, it includes at least the parameter R characterizing the distribution of the neutron flux in the core, for example the axial offset AO.
• the characteristic quantities of the state of the core calculated by the predictive model 71 are those described above for the control method.
• the cost function calculated by the module 73 characterizes for example a change in a difference between said parameter R characterizing the distribution of the neutron flux in the core and a reference value over the determined time interval.
• the cost function is typically as described above for the steering method. Other cost functions can be considered, as described below.
• the module 73 uses to calculate the cost function the evolution calculated by the predictive model 71.
• the optimization algorithm 75 is programmed to, iteratively, generate an injection sequence, have the evolution of the at least one corresponding characteristic quantity calculated by the predictive model of the heart 71, have the cost function evaluated corresponding by the cost module 73, until a convergence criterion of the cost function is satisfied.
• the convergence criterion includes for example reaching the extremum of the cost function.
• this extremum is a minimum.
• the convergence criterion may provide that one or more of the following constraints are satisfied: - the difference between the parameter R characterizing the distribution of the neutron flux in the core and the reference value during said determined time interval remains constantly below a determined limit;
• the convergence criterion, in particular the constraints, are as described above for the piloting process.
• the calculation unit 69 also includes a slope module 77 programmed to determine an optimum slope for the evolution of the reactor power as a function of time during the power variation from the first power to the second power.
• a slope module 77 programmed to determine an optimum slope for the evolution of the reactor power as a function of time during the power variation from the first power to the second power.
• the slope module 77 is programmed for:
• the predictive model of the heart 71 calculate the evolution of at least one quantity characteristic of the state of the heart of the nuclear reactor during said power variation, for several slope values, the injection of neutron poison or d water per unit of time being considered constantly equal to the maximum possible;
• the slope retained by the predictive model 71 to determine the optimum injection sequence is the optimum slope, minus possibly a safety factor. It is also recommended for the operator.
• This optimum slope is the fastest slope that can be obtained taking into account the existing constraints for the reactor.
• the calculation unit 69 does not include the slope module 77.
• the slope to be used by the predictive model 71 is then fixed by the operator, and entered using the user interface 65.
• the slope module 77 preferably determines the optimum slope as described for the steering method.
• Set 63 is also configured to display:
• the optimum neutron poison and/or water injection sequence i.e. the one for which the convergence criterion of the cost function has been reached;
• This information is typically displayed on user interface 65.
• Figure 4 shows an example screen of user interface 65 at time t0.
• the upper part of the screen indicates the injection instructions.
• the injection instructions located above the horizontal line are neutron poison injection instructions, those below the horizontal line are water injection instructions.
• the lower part of figure 4 illustrates the evolution of one of the characteristic parameters of the state of the heart, as a function of time.
• One or more parameters can be displayed. As indicated above, this or these parameters are chosen from among the parameter R characterizing the distribution of the neutron flux in the core, the average temperature of the primary coolant fluid Tmoy, the power of the core P, the position of the control groups Pbank, the concentration xenon Xe, or the concentration of neutron poison in the primary coolant Cpn.
• FIG. 4 illustrates a situation where a new injection sequence is determined with the assembly 63 every ten minutes, this sequence comprising six injection operations lasting ten minutes each, therefore covering a time interval of one o'clock.
• the upper part of the screen shows the optimum injection sequence calculated using the operating parameters acquired at t0.
• the lower part of the screen shows the evolution of the characteristic parameters of the state of the heart calculated for the optimum injection sequence.
• the figure shows the injection calculated at the previous iteration, i.e. calculated on the basis of the operating parameters acquired at t0-10 min.
• Figure 4 shows between t0-10 min and t0 the injection actually performed by the operator, and the evolution of the parameter characterizing the state of the heart actually measured, using instrumentation 55.
• step S10 of the control method is carried out manually by the operator, who enters the power program on the user interface 65.
• Step S20 is performed by acquisition unit 67.
• Step S30 is performed by the calculation unit 69.
• Sub-step S31 is performed by optimization algorithm 75.
• Sub-step S32 is executed by the predictive heart model 71.
• Sub-step S33 is executed by cost function 73.
• Sub-step S34 is executed by optimization algorithm 75.
• Sub-step S35 is executed by slope module 77.
• Step S40 is typically performed on user interface 65.
• control unit 63 provides the operator with the optimum neutron poison and/or water injection sequence determined for the time interval considered, possibly accompanied by the slope retained to carry out the power variation.
• the operator drives units 31 and 39 directly, depending on the optimum injection sequence provided by unit 63.
• valves 37 and 47 controls valves 37 and 47, and pumps 35 and 45.
• the cost function is different from that described above.
• the predictive model of the core does not take into account any dead band around the reference temperature Tref.
• a displacement of the groups of control bars is determined as soon as Tmoy deviates from Tref.

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• 2021
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