Device for detecting a concentration of neutron absorber in a primary circuit.

Abstract

Device (1) for measuring a concentration of a neutron absorber in a fluid (3), the device being intended to be disposed facing a pipe (2) of the fluid, the device comprising: a cavity (10), delimited by a shell (11), the cavity having a front end, intended to be disposed facing the pipe; a neutron source (13), extending between the shell and the front end; at least one neutron detector (14), disposed in the cavity, generating a detection signal dependent on the irradiation of the cavity by the neutrons; the device being characterized in that it comprises a neutron reflector (15), the neutron source being disposed between the front end and the neutron reflector, the neutron reflector being configured to reflect neutrons, emitted by the neutron source, back to the front end of the cavity.

Description

Title of the invention: Device for detecting a concentration of neutron absorber in a primary circuit. technical field

[0001] The technical field of the invention is the measurement of a neutron absorber in a fluid, one intended application being the determination of the concentration of boron (or other absorbing isotope) in the water of the primary circuit of a nuclear reactor. PRIOR ART

[0002] In a water-cooled nuclear reactor, reactivity must be controlled to prevent the occurrence of a criticality situation. Reactivity in the primary circuit is adjusted by adding an isotope with high neutron absorption. Such an absorbing isotope can be 10B, added to the water in the form of boric acid (H3BO3).

[0003] In pressurized water reactors, the boron concentration in the primary circuit water is adjusted by the chemical and volume control system, usually referred to by the acronym RCV, or the Anglo-Saxon designation CVCS (Chemical and Volume Control System). The boron concentration can be increased or decreased by adding borated water (water containing boric acid) or demineralized water. The RCV is an auxiliary circuit of the primary circuit.

[0004] The boron concentration in the primary circuit is an important parameter for controlling nuclear reactors. For this purpose, boron concentration measurement devices, known as "boron meters" or BCMS (Boron Concentration Measuring Systems), are used. These devices are based on the absorption of neutrons by boron. A neutron source irradiates a pipe, generally in the RCV circuit, with a diameter of approximately 10 cm. A neutron detector is positioned near a pipe in the RCV circuit. The detector is positioned so as to be exposed to neutrons emitted by the source, which have passed through the pipe before interacting with the detector. The higher the boron concentration in the primary circuit, the greater the neutron absorption in the primary circuit, and the lower the number of neutrons detected by the detector.An example of sizing a BCMS type system was described in the publication Nouhaila Tabti, Adrien Sari, and Dominique Tromson, Toward the enhancement of nuclear reactor safety: physical phenomena involved in the measurement accuracy of boron concentration online monitoring Systems. IEEE. Transactions on Nuclear Science, vol. 71, no. 5, pp. 1010-1018, 2024, doi: 10.1109 / TNS.2024.3351481.

[0005] The publication Tabti N. et al., "Exploring Boron monitoring solutions in pressurized water reactors for enhanced loss of coolant accidents mitigation," describes the possibility of installing a boron measurement system around a pipe in the primary circuit of a pressurized water reactor. Indeed, the ability to measure the temporal evolution of boron concentration within the primary circuit can be of interest, particularly during an accident causing a loss of water from the primary circuit. This type of accident is usually referred to as a "Loss of Coolant Accident (LOCA)."

[0006] To this end, the inventors have designed a device allowing monitoring of the concentration of boron, or other neutron absorber diluted in a heat transfer fluid, in the primary circuit of a reactor. Description of the invention

[0007] A first object of the invention is a device for measuring the concentration of a neutron absorber in a fluid, the device being intended to be positioned opposite a fluid conduit, the device comprising: - a cavity, delimited by a shell, the cavity having a front end, intended to be positioned facing the pipe; - a neutron source, extending between the hull and the forward end; - at least one neutron detector, located in the cavity, generating a detection signal dependent on the irradiation of the cavity by neutrons; - a processing unit, configured to estimate the concentration of the neutron absorber in the fluid from the detection signal; - the device being characterized in that it comprises a neutron reflector, the neutron source being disposed between the front end and the neutron reflector, the neutron reflector being configured to reflect neutrons, emitted by the neutron source, back to the front end of the cavity.

[0008] In other words, the neutron reflector is configured to reflect neutrons back towards the duct by backscattering. This increases the number of neutrons emitted by the source and propagating inside the duct.

[0009] The neutron source is disposed in the cavity. The cavity can be hollow, i.e. filled with air, or filled with a material suitable for neutron backscattering, forming a reflector.

[0010] According to one possibility, the device includes a front screen, disposed between the neutron detector or detectors and the front end, and configured to attenuate at least 75% of gamma radiation with an energy of 1 MeV emitted by the conduit.

[0011] The front screen can be positioned at the front end.

[0012] The cavity can be delimited on one side by the shell, on the other side by the front screen.

[0013] The neutron source can be surrounded by the front screen.

[0014] The cavity may have an opening at the front end.

[0015] The neutron reflector is preferably made of a material whose albedo is greater than 0.7, the albedo being a ratio between a neutron flux incident on the material and a neutron flux backscattered by the material.

[0016] The neutron reflector is for example made of a material comprising carbon and / or oxygen and / or hydrogen and / or beryllium.

[0017] The neutron reflector may comprise a material selected from MgO, Graphite, Zr3Si2, MgAl2O4, Y3A15O12, SiC, ZrC, TiC, V-4Ti-4Cr, steel, UO2.

[0018] According to one possibility, the neutron reflector extends to a distance from the neutron source of less than 20 cm.

[0019] According to one possibility, the cavity extends around a center, positioned at an equal distance between the hull and the front end, the center of the cavity being located less than 20 cm from the front end.

[0020] The neutron reflector can extend, in or along the hull, with a thickness greater than 2 cm and less than 20 cm.

[0021] According to one possibility, the neutron detector or each neutron detector extends between the neutron source and the shell.

[0022] The reflector can extend into the hull or along the hull.

[0023] According to one possibility, the device includes at least one auxiliary detector configured to generate an auxiliary detection signal dependent on the gamma irradiation level, the processing unit being configured to: - estimate a gamma irradiation level from the auxiliary detection signal; - from the gamma irradiation level, determine a component of the detection signal, resulting from the detector(s), relating to the gamma irradiation; - from said component, extract a component of the detection signal, resulting from the detector or each detector, relating to neutron irradiation.

[0024] The invention will be better understood upon reading the description of the exemplary embodiments presented later in this description, in connection with the figures listed below. FIGURES

[0025] Figure 1 schematically illustrates the main elements of a primary circuit of a pressurized water reactor.

[0026] Figure 2 shows a first embodiment of a device according to the invention.

[0027] Figures 3A and 3B illustrate two detector arrangements of the device, the detectors being respectively separated from each other according to two different angle values.

[0028] Fig. 3C represents the reaction rate (n,a) as a function of the value of the angle between the two detectors.

[0029] Fig. 3D shows the resolving power of a measurement as a function of the value of the angle between the two detectors.

[0030] Fig. 4A shows the evolution of the reaction rate (n, a) as a function of the boron concentration for different temperatures of the heat transfer fluid (Tf) and the detectors (Td).

[0031] Fig. 4B represents the evolution of the resolving power as a function of the boron concentration for the different configurations described in relation to Fig. 4A.

[0032] Fig. 5 shows the evolution of thermal or neutron sensitivities as a function of boron concentration in the heat transfer fluid.

[0033] Fig. 6A shows the evolution of the reaction rate as a function of boron concentration, using different types of source.

[0034] Fig. 6B represents the evolution of the resolving power as a function of the boron concentration for the different configurations described in relation to Fig. 6A.

[0035] Figure [Fig.7] illustrates a backscattering of neutrons by a reflector.

[0036] Fig. 8A shows the evolution of the reaction rate as a function of boron concentration, in the absence of a reflector or in the presence of different types of reflectors.

[0037] Fig. 8B represents the evolution of the resolving power as a function of the boron concentration in the different configurations described in relation to Fig. 8A.

[0038] Figures 9A, 9B and 9C illustrate three device configurations according to the invention.

[0039] Fig. 9D shows the evolution of the reaction rate (n, a), as a function of the boron concentration, for the different configurations shown schematically in figures 9A to 9C.

[0040] Fig. 9E shows the evolution of the resolving power, as a function of the boron concentration, for the different configurations shown schematically in Figures 9A to 9C.

[0041] Figures 10A and 10B illustrate two device configurations according to the invention.

[0042] Fig. 10C shows the evolution of the reaction rate (n, a), by changing the position of the center of the cavity relative to the conduit, as shown in Figures 10A and 10B.

[0043] Fig. 10D shows the evolution of the resolving power as a function of changing the position of the center of the cavity relative to the conduit, as shown in Figures 10A and 10B.

[0044] Figures 11A and 11IB illustrate two device configurations according to the invention.

[0045] Fig. 1 IC shows the evolution of the reaction rate (n, a), as a function of the position of the center of the cavity relative to the conduit, for configurations such as those shown schematically in Figures 11A and 11B.

[0046] Figures 12A and 12B illustrate two device configurations according to the invention.

[0047] Fig. 12C shows the evolution of the reaction rate (n, a), as a function of the radius of the cavity, for configurations such as those shown schematically in Figures 12A and 12B.

[0048] Fig. 12D shows the evolution of the resolving power, as a function of the cavity radius, for configurations such as those schematically represented in Figures 12A and 12B.

[0049] Fig. 12E shows the evolution of the acquisition time, as a function of the thickness of an MgO reflector, to obtain a measurement whose accuracy is less than or equal to a predetermined threshold.

[0050] Figure 13 shows a second embodiment of a device according to the invention.

[0051] Figures 14A, 14B, 14C and 14D illustrate four measurement configurations. Figures 14A and 14B correspond to the first embodiment. Figures 14C and 14D correspond to the second embodiment.

[0052] Fig. 14E shows the evolution of the reaction rate (n, a), as a function of boron concentration, for the configurations shown schematically in Figures 14A to 14D.

[0053] Fig. 14F shows the evolution of the resolving power, as a function of boron concentration, for the configurations shown schematically in Figures 14A to 14D.

[0054] Fig. 15 shows the evolution of thermal or neutron sensitivities as a function of the concentration of boron in the heat transfer fluid for a so-called reference configuration.

[0055] Figures 16A and 16B show a third embodiment of a device according to the invention. PRESENTATION OF SPECIFIC IMPLEMENTATION METHODS

[0056] Figure 1 schematically illustrates the main elements of the primary circuit of a pressurized water reactor. In this example, the reactor comprises a vessel V connected to several cooling loops, forming the primary circuit through which a heat transfer fluid circulates. In this example, the heat transfer fluid 3 is pressurized water: the water is maintained under pressure, for example at 155 bar, of so as to remain liquid up to a temperature above 300 °C. Each cooling loop comprises a hot section H and an intermediate section I on either side of a steam generator GV. The intermediate section I extends to a pump P. The intermediate section I includes a U-shaped branch extending between the steam generator GV and the pump P. The cold section extends between the pump P and the reactor vessel V. The primary circuit consists of pipes 2, extending between several components such as steam generators and pumps; some components are not shown (pressurizer, injection circuits, etc.). In this type of reactor, the concentration of 10B generally varies between 1000 ppm and 3000 ppm depending on whether the reactor is operating or shut down. In this presentation, the concentration of 10B is assumed to vary between 0 and 5000 ppm.

[0057] Figure 2 represents an example of a first embodiment of a device 1 according to the invention. The device 1 is arranged to be positioned opposite a conduit 2 of the primary circuit, and preferably in contact with the primary circuit. The device may in particular be positioned at the U-shaped branch.

[0058] Device 1 is intended to estimate the concentration of an isotope 4 in the heat transfer fluid 3. Isotope 4 is a neutron-absorbing isotope. In this example, the isotope is 10B. In this example, the heat transfer fluid 3 is borated water (a mixture of water and boric acid). The diameter of the pipe 2 is typically between 80 and 100 cm. The device is positioned facing the pipe 2, and preferably in contact with it.

[0059] Placing the device 1 opposite the primary circuit has several advantages, compared to the prior art configuration, according to which a boron measurement system is placed around a conduit of the RCV circuit: - a measurement of boron concentration carried out in real time, directly on the primary circuit, and not on an auxiliary circuit such as the RCV circuit. This provides real-time knowledge of the neutron-absorbing content of the heat transfer fluid, including in the event of unavailability of the RCV circuit, for example following the closure of a valve of said RCV circuit; - faster detection of potential malfunctions, such as a boron dilution accident, whether homogeneous or heterogeneous, caused by an anomaly that could lead to operation outside the safety framework. An example of such an anomaly is a primary refrigerant loss-of-coolant accident (PRLA) inducing a heterogeneous dilution phenomenon inherent to the accidental transient.

[0060] However, being coupled to the primary circuit presents technical difficulties, including the large size of the primary circuit pipes, generally between 80 cm and 100 cm. These diameters are much larger than those of the RCV circuit pipes, which are generally less than 10 cm.

[0061] In addition to this, there is potentially significant γ radiation due to the presence of γ-emitting isotopes in the primary circuit: these isotopes correspond to activation products present in the heat transfer fluid (for example 60Co), to which fission products (for example 137Cs) may potentially be added.

[0062] Another constraint is a potential temperature variation: In a pressurized water reactor, during shutdown phases, the temperature of the water in the primary circuit can be around 50°C. During operation, the water temperature can reach or exceed 280°C (at the core inlet) and 330°C (at the core outlet).

[0063] The device comprises a cavity 10, which, in this example, is filled with air. The cavity 10 extends between a shell 11 and a front end 12. The shell 11 extends opposite the front end 12. The front end 12 is intended to be positioned facing the conduit 2. In this example, the cavity 10 has a front opening, at the front end 12, leading to the conduit 2. In this example, the cavity is substantially hemispherical in shape. The cavity can extend over a diameter (or longer diagonal) of between 10 cm and 30 cm, and preferably between 10 cm and 20 cm.

[0064] The device includes a neutron source 13. This can be an isotopic source, comprising one or more isotopes, enabling the emission of neutrons. In this example, the isotopic source is an Am-Be type source, whose neutron emission is based on an (a, n) type reaction. Other types of sources, based on the same principle, are conceivable, for example, 241Am-Li or 244Cm-Be. The neutron source 13 can be a spontaneous fission type source, the most common isotopes being Cf or Cm. It can also be a neutron generator. The use of a neutron generator allows precise control of the neutron emission, at the expense of significant size and safety-related difficulties for use within a reactor building. In the following description, the neutron source used is an Am-Be type source, unless otherwise specified.

[0065] The device comprises at least one detector 14, extending between the shell 10 and the front end 12. In the example shown, the device comprises two detectors 14, extending into the cavity 10. Using two detectors provides redundancy in measurements. This allows for the detection of a malfunction. affecting one of the two detectors. The device may have more than 2 detectors.

[0066] Each detector 14 is configured to provide a counting rate TC corresponding to a number of neutrons detected per second. In this example, each detector 14 is a boron-deposited gaseous proportional counter. This type of counter is common. Under the effect of irradiation by a neutron flux, charged particles (α or βLi) are formed by (α,β) capture. The charged particles are detected by electrodes polarized by secondary ionizations in the detector. In the following examples, each neutron detector is a Photonis CPNB64 type proportional counter, whose detection efficiency is 25 counts / cm² (or counts per second per neutron per cm² per second).

[0067] Each detector 14 is connected to a processing unit 20 by a wired connection, preferably robust against neutron and photon radiation. The processing unit 20 may include a microprocessor. The processing unit 20 is programmed to perform an estimation of a concentration of 10B in the fluid 3 based on the count rates TC measured by each detector 14.

[0068] The processing unit 20 is programmed to calculate a count rate TC, resulting from a detector 14. The presence of different detectors introduces redundancy, which makes it possible to identify any malfunction of one of the detectors. From the count rate, the concentration of 10B is estimated, taking into account the temperature Tf of the water in the primary circuit. For example, a calibration base has been previously stored in the processing unit. The calibration base includes, for different values ​​of the concentration of 10B, and for different values ​​of the temperature T, values ​​of the count rate TC. The water temperature Tf can be obtained by a thermocouple dedicated to the device 1 or by external measuring means.

[0069] Some analytical expressions are proposed in the literature that establish a relationship between boron concentration and count rate. For example, the publication Pirat P, "Boronline, a new generation of boron meter," 2nd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and their Applications, 2011, IEEE, pp. 1-2, describes an expression of the type:

[0070] TC =____r1_____( 1 ) a(CB) +bCB+c

[0071] Where a, b and c are adjustment parameters: these are real numbers which depend on the temperature of the heat transfer fluid, the detector used and the configuration in which the detector is located, taking into account for example the presence as well as the geometry and nature of a moderator material in the vicinity of the detector.

[0072] A Shinkori equation, well-suited to high boron concentrations, can also be implemented. This equation comes from the publication by Si Hyeong Sung and Hee Reyoung Kim, “Experimental characterization of the accuracy of multidetector boron meter for operational safety of reactors.” International Journal of Energy Research, 42(8):2701-2709, 2018. According to this equation:

[0073] r - o + (l')ou

[0074] T c = aL+a2CB (1 ”) l+a3CB+a4(CB)

[0075] The parameters ^q, âp Ô2 and a3 are real numbers which depend on the temperature of the heat transfer fluid, the detector used and the configuration in which the detector is located.

[0076] Different arrangements for the detectors 14 in the cavity 10 were simulated for the same position of the neutron source 13 and by varying an angle 0 between two segments connecting each detector to the center of the conduit 2. The conduit 2 extended along internal and external radii respectively equal to 35 cm and 41 cm, i.e. a conduit with an external diameter of 82 cm and a thickness of 6 cm.

[0077] Figure 3A corresponds to a transmission configuration, in which the value of angle θ is greater than 180°. In this configuration, the detectors are exposed to a neutron flux that has passed through the pipe, undergoing attenuation due to the presence of ¹⁰B in the coolant. Figure 3B corresponds to a backscattering configuration, in which the value of angle θ is less than 180°. In this configuration, the detectors are exposed to a neutron flux that has been backscattered by the coolant.

[0078] The counting rate was estimated for different values ​​of the θ angle. To do this, the reaction rate r(n,a) was calculated, which corresponds to the number of neutron capture reactions in the Photonis CPNB64 boron-deposited proportional counter detector, per neutrons emitted by a 2.5 MeV neutron source. Figure 3C shows different values ​​of r(n,a) for different values ​​of the θ angle, with a 10B concentration in the coolant of 0 ppm. It can be observed that the reaction rate is much higher in backscattering configurations, particularly for θ angles less than 90°, and especially for θ angles less than 30°.

[0079] For different angles 0, the CR was also determined, see [Fig. 3D], the power resolution being defined by:

[0080] TC Oppni TCxppm TCoppm T Cxppm TCoppm (2)

[0081] Where TCQppm and TCxppm are the counting rates when the concentration of 10B is respectively equal to 0 ppm and x ppm. For each angle 0, the resolving power R was calculated with x = 5000 ppm.

[0082] Each count rate corresponds, at least to a first order, to the reaction rate multiplied by the detector efficiency and by the emission intensity of the neutron source. Thus, the count rate is proportional to the reaction rate.

[0083] It is observed that when 0 < 90°, the resolving power R decreases. However, for these angles, the relative decrease in resolving power is less than the increase in the reaction rate. It is therefore preferable that the detectors be arranged on either side of the source, preferably equidistant from it, and be angularly separated by an angle 0 < 90°, and preferably 0 < 30°, from the center of the pipe.

[0084] The reaction rate r(n,a), as previously defined, was estimated for different temperatures Tf of the primary circuit water and for different detector temperatures Td. The detector temperature Td corresponds to the ambient temperature, i.e., the temperature of the medium 5 surrounding pipe 2. Figure 4A shows the reaction rate r(n,a) (ordinate axis) for different concentrations of 10B (abscissa axis). Figure 4B shows the resolving power R, taking into account x = 5000 ppm (ordinate axis), defined in (2), for each concentration of 10B (abscissa axis). In Figures 4A and 4B, three configurations were tested: Tf = 277°C and Td = 327°C; Tf= 277°C and Td= 20°C; Tf= 20°C and Td= 20°C.

[0085] It is observed that the variation in ambient temperature has little effect on the reaction rate or the resolving power. However, the neutron performance of the measurement is influenced by the temperature Tf of the heat transfer fluid: the higher the temperature of the heat transfer fluid, the lower the capture cross-section, as does the density of the water, which increases the quantity of backscattered neutrons, and therefore the quantity of neutrons detected. From the modeled reaction rate values ​​r(n,a), the thermal sensitivity and neutron sensitivity were estimated at 277°C and 20°C.

[0086] Thermal sensitivity is defined by Stb = (3). It quantifies the variation of the counting rate, at a constant concentration of 10B, under the effect of a variation in temperature. Neutron sensitivity is defined by = Ï|LL (4). It quantifies the variation of the counting rate, at constant temperature, under the effect of a concentration variation of 10B.

[0087] Figure 5 shows the thermal sensitivity (curve a - unit: s'.°C') and the neutron sensitivities for heat transfer fluid temperatures of 277°C (curve b) and 20°C (curve c). The unit for neutron sensitivities is s'.ppm'. This means that the device's response is more influenced by a change in the heat transfer fluid temperature than by a change in the concentration of 10B. Therefore, it is useful to have a value for the heat transfer fluid temperature to obtain a reliable estimate of the 10B concentration over the concentration range of 0 to 5000 ppm. Neutron source

[0088] The use of three different neutron sources has been modeled: an Am-Be source, a DD (Deuterium-Deuterium) type neutron generator, producing neutrons with an energy of 2.5 MeV, and a DT (Deuterium-Tritium) type neutron generator, producing neutrons with an energy of 14 MeV.

[0089] Figure 6A shows the reaction rate r(n,a) (ordinate axis) for different concentrations of 10B (abscissa axis). Figure 6B shows the resolving power R (ordinate axis), defined in (2), for each concentration of 10B (abscissa axis). In Figures 6A and 6B, curves a), b), and c) correspond respectively to the Am-Be source and to neutron generators with emission energies of 2.5 MeV and 14 MeV. It can be observed that the neutron detection performance is quite similar for the Am-Be source and the DD-type neutron generator. Hull

[0090] The shell 11 delimiting the cavity 10 may advantageously include shielding 16 so as to form a screen for each detector 14 against ambient gamma radiation. The shielding 16 is preferably made of a metal with an atomic number Z greater than 60, or even 80, and a high density, preferably greater than 10 g / cm³. It may, in particular, be lead or a tungsten alloy. The thickness of the shielding may, for example, be 10 cm. It is preferably between 2 cm and 20 cm. A 2 cm thickness of lead allows the absorption of more than 75% of gamma radiation with an energy of 1 MeV. Increasing the thickness may result in an excessive mass of the shell. A thin layer of cadmium 17, with a thickness between 0.5 mm and 2 mm, can cover the shielding 16, for reasons of radiation protection, cadmium having a high absorption of neutrons.

[0091] An important aspect of the invention is the use of a reflector 15. By reflector, we mean a component that allows for the backscattering of neutrons so as to maximize the number of neutrons propagating in the duct. The backscattered neutrons are Neutrons initially emitted in a direction opposite to the pipe then undergo one or more successive scatterings in the reflector, so as to propagate towards the pipe. The backscattering power can be characterized by an albedo [3], which corresponds to a ratio between the flux J passing through a surface S, in the opposite direction to a flux incident on said surface

[0092] Thus, _ J (5)

[0093] The albedo of a reflector can be easily estimated, by modeling, as a function of the material forming the reflector and its thickness. The modeling can be performed using a particle transport code of the MCNP type.

[0094] It is preferable that the characteristics of the reflector (material, thickness) be such that the albedo (with respect to neutrons) is greater than 0.7, or even greater than 0.8. It is often considered that a good neutron reflector has an albedo between 0.7 and 0.9. Figure 7 schematically illustrates the backscattering of a neutron emitted by a reflector 15. In this figure, the surface S, forming the interface between the cavity 10 and the reflector 15, has been considered to be planar.

[0095] The reflector 15 can extend to a thickness of between 2 cm and 20 cm, or between 2 cm and 15 cm. Various materials suitable for forming a reflector were tested: - Graphite: albedo 0.937 - density 1.6 g.cm3; - MgO: magnesium oxide: albedo 0.859 - density 3.58 g.cm3; - Zirconium silicide (Zr3Si2): albedo 0.833 - density 5.88 g.cm3.

[0096] Other materials can form a reflector, in particular materials containing light atoms, which are conducive to elastic neutron scattering, for example C or O or H or Beryllium. The reflector can be formed from a polymer, for example polyethylene.

[0097] In addition to the materials mentioned above, the reflector can be made of at least one of the materials chosen from: MgAl2O4, Y3A150i2, SIC (silicon carbide), ZrC, TiC, or V-4Ti-4Cr type alloys, steel, or even UO2 (with depleted U).

[0098] The use of three different reflectors, 5 cm thick respectively made of Graphite, MgO and Zr3Si2, was modeled. During the modeling, a Pb shielding thickness of 3.5 cm was taken into account.

[0099] Figure 8A shows the reaction rate r(n,a) (ordinate axis) for different concentrations of 10B (x-axis). Figure 8B shows the resolving power R (y-axis) defined by equation (2) for each concentration of 10B (x-axis). In Figures 8A and 8B, curves a), b), and c) correspond respectively to graphite, MgO, and Zr3Si2 reflectors. A reflectorless configuration was also modeled: curve d). It can be observed that the use of A reflector leads to a significant increase in the reaction rate r(n,a), while moderately reducing the resolving power R. Thus, the use of graphite or MgO type reflectors results in an increase in the reaction rate of approximately a factor of 6. The use of a Zr3Si2 type reflector results in an increase in the reaction rate of approximately a factor of 4.

[0100] The best compromise between increasing the reaction rate and decreasing the resolving power is obtained with an MgO type reflector.

[0101] Three different configurations were modeled, combining a Pb shield 16, 3.5 cm thick, and a reflector 15, 5 cm thick, made of MgO. In these three configurations, the positions of the neutron source and the detectors in the cavity were modified. - Configuration a): Detectors 14 centered in cavity 10, with the neutron source 13 positioned between the detectors and the shell 11: see [Fig. 9A] - Configuration b): Detectors 14 centered in cavity 10 and positioned between the neutron source 13 and the shell 11: see [Fig. 9B] - configuration c): detectors 14 offset in cavity 10, being placed near the front end 12 so that the detectors and the source 13 are positioned against the pipe, or at a short distance (5 cm or 2 cm or 1 cm) from it: cf. [Fig.9C].

[0102] Figure 9D shows the reaction rate r(n,a) (ordinate axis) for different concentrations of 10B (abscissa axis). Figure 9E shows the resolving power R (ordinate axis), defined in (2), for each concentration of 10B (abscissa axis). It can be observed that the most favorable configurations are those in which the neutron source 13 is arranged between the detectors 14 and the conduit 2 (Figure 9C). It is optimal to place the detectors 14 and the source 13 as close as possible to the conduit. Thus, the detectors 14 are preferably arranged as close as possible, even at the front end 12 of the cavity, for example at a distance of less than 5 cm, 2 cm, or 1 cm from the front end 12 of the cavity. The distance of detectors 14 and neutron source 13, relative to conduit 2, induces a degradation of reaction rate and resolving power.It may be useful to slightly offset the detectors and / or neutron source from the pipe to avoid exposure to excessively high temperatures or vibrations. In a pressurized water reactor, the pipe temperature can reach 300°C.

[0103] Based on the configuration of [Fig. 9C], the position of the center O of the cavity relative to the front end 12 of the cavity, in contact with the pipe, was modified. In these models, the interface between the cavity and the shell was considered to follow A circular contour with a diameter of 28 cm was modeled. Distances between the center of cavity O and the center of the duct were modeled in 2 cm increments, ranging from 36 cm (see Fig. 1OA) to 54 cm (see Fig. 1OB). This is equivalent to varying the thickness of the material filling the cavity, in this example air, behind the detectors. In the configuration shown in Fig. 1OA, the center of cavity O is offset 5 cm inward from the distal end 12 of the duct.

[0104] Fig. 1OC shows the reaction rate r(n,a) (ordinate axis) for different distances (abscissa axis). Fig. 1OD shows the resolving power R (ordinate axis), defined in (2), taking x = 5000 ppm, for each distance (abscissa axis). It can be observed that the reaction rate decreases with distance, while the resolving power remains relatively stable. It is therefore preferable to maintain the distance between the detectors and the front end 12 as small as possible, with the source positioned between the detectors and the front end, so as to be as close as possible to the pipe.

[0105] Based on the configuration of [Fig. 1OA], the cavity radius was varied between 10 cm ([Fig. 11A]) and 20 cm ([Fig. 1IB]), in increments of 2 cm. [Fig. 1IC] shows the reaction rate r(n,a) (ordinate axis) for different radii (abscissa axis). It can be observed that the influence of the cavity radius on the reaction rate is small. Subsequently, a cavity with a radius of 12 cm was considered.

[0106] Considering a cavity with a radius of 12 cm, filled with air, and detectors positioned at the front end of the cavity, thus in contact with the pipe, the effect of varying the thickness of an MgO reflector on detection performance was modeled. A reflector thickness varying between 2 cm (see [Fig. 12A]) and 16 cm (see [Fig. 12B]), in 2 cm increments, was modeled. [Fig. 12C] shows the reaction rate r(n,a) (ordinate axis) for different thicknesses (abscissa axis). [Fig. 12D] shows the resolving power R (ordinate axis), defined by equation (2), for each thickness, taking x = 5000 ppm. In the configurations shown in figures 12A and 12B, the thickness of the lead shielding 16 is 3.5 cm.

[0107] For each thickness, an acquisition time A t was estimated to obtain a quantification accuracy of 10B of 2% (to ± 20), using an Am-Be type neutron source whose emission intensity is 107 ns '.

[0108] The uncertainty dTC on the counting rate can be determined, from the acquisition time A t, by the equation [°109] dTC = k^(6)

[0110] k being a predetermined positive real number, for example k = 2 if one wants a precision of 20. [YES] From the dTC values, using (1”), we calculate dCp for the different acquisition periods △ t considered. We then look for the value of A t for Muette ^ 002

[0112] It is understood that the value of 0.02 corresponds to an arbitrary relative error, which can be decreased (increasing the value of A t) or reduced, depending on the operating specification of the nuclear reactor.

[0113] The TC counting rate is estimated from r(n,a), multiplied by the emission intensity of the source and the efficiency of the detector.

[0114] Figure 12E shows the acquisition time as a function of the reflector thickness. Formula (1”) was implemented with β1 = 1647.252, β2 = 0.928, β3 = 8.987.104, β4 = 1.103.108. For a reflector made of MgO, a thickness of 8 cm is a good compromise between the sensitivity of the measurement, which depends on the reaction rate, and the resolving power. This results in the shortest possible acquisition time to obtain the desired accuracy. More generally, it should be noted that the reflector thickness is preferably greater than 5 cm, and for example between 5 cm and 15 cm or 20 cm, bearing in mind that the optimal thickness depends on the material forming the reflector.

[0115] Figure 13 illustrates a second embodiment in which the device comprises a front screen 18 at the front end 12. The front screen 18 is made of a material similar to the material forming the shielding 16, that is, a material with a sufficiently high atomic number and density, for example, lead or a tungsten alloy, to satisfactorily attenuate (at least 75%, or even 90%, or more) high-energy gamma radiation with an energy of 1 MeV. The front screen 18 is intended to protect the detectors 14 from the radiation emanating from the conduit. Thus, the front screen 18 is interposed either at the front end 12 or between the latter and each detector.

[0116] Four different configurations were modeled. In two configurations, shown in Figures 14A and 14B, the shell is surrounded by a Pb shield 16, with respective thicknesses of 5 cm and 10 cm. In two other configurations, shown in Figures 14C and 14D, the cavity 10 is delimited by a front screen 18, interposed between the detectors 14 and the front end 12. The shell has a Pb shield 16 with a thickness of 10 cm. The front Pb screen 18 has a maximum thickness of 9 cm, the maximum thickness of the screen being located at the front end 12, perpendicular to the conduit.

[0117] In the configurations shown in Figures 14A, 14B, the cavity is filled with air and the shell includes an 8 cm thick MgO reflector. The maximum thickness of the cavity 10 is approximately 7 cm.

[0118] In the configuration shown in Figures 14C, the cavity is filled with air and the shell 11 has a 5 cm thick MgO reflector. The maximum thickness of the cavity 10 is 8 cm.

[0119] In the configuration shown in [Fig. 14D], the cavity 10, in which the detectors 14 are located, is formed of the MgO reflective material, with a maximum thickness of 8 cm. The detectors are therefore embedded in the material forming the reflector. The shell 11 then consists solely of the shielding 16.

[0120] In the configurations shown in Figures 14C and 14D, the neutron source 13 is arranged within the front screen 18, at the front end 12. It is placed in a channel provided in the front screen 18.

[0121] Figure 14E shows the reaction rate r(n,a) (ordinate axis) for different concentrations of 10B (abscissa axis). Figure 14F shows the resolving power R (ordinate axis), defined in (1), for each concentration of 10B (abscissa axis). In Figures 14E and 14F, the configurations of Figures 14A, 14B, 14C, and 14D are respectively identified by curves a), b), c), and d). Curve e) corresponds to the reference configuration as described in connection with Figures 12A and 12B, with a reflector thickness of 8 cm, a lead shielding thickness of 3.5 cm, and a maximum air cavity thickness of 7 cm.

[0122] It is observed that increasing the thickness of the shielding 16 results in a decrease in the resolving power R and an increase in the reaction rate. The shielding 16 increases the quantity of neutrons emitted by the source and then backscattered towards the conduit.

[0123] The configuration shown in [Fig. 14C] results in a slight decrease in the response rate, although this remains close to the reference configuration. Thus, the presence of the front screen 18, of moderate thickness, typically less than 10 cm, does not appear to have a detrimental effect on the response rate, and therefore on the sensitivity of the device. However, the presence of the front screen reduces the resolving power.

[0124] The configuration shown in [Fig. 14D] is accompanied by a significant increase in the reaction rate, on the order of a factor of 3, compared to the reference configuration. However, this is accompanied by a significant reduction in resolving power.

[0125] For each of the configurations described in relation to Figures 14A to 14D, the acquisition time △ t was estimated to obtain an accuracy of 2% for the estimation of the concentration of 10B.

[0126] Table 1 summarizes the values ​​of the parameters ai, a2, ^4 for each configurations, as well as the acquisition time At normalized by the acquisition time Aq which corresponds to the reference configuration. A = 495 seconds. The acquisition time A t can be reduced by increasing the emission intensity of the neutron source or by decreasing the uncertainty. configuration 14A 14B 14C 14D ai 1985.984 2610.3627 1421.307 4108.3099 a2 1.094 1.495 1.053 3.072 a3 8.634 104 8.313 104 1.044 10 4 8.531 104 *4 6.710 10 9 4.766 10 9 1.286 10 8 2.407 10 9 At At0 1.13 1.26 1.67 4.63

[0127] Table 1

[0128] The configurations shown in Figures 14A and 14B exhibit the detection performance closest to the reference configuration. However, the configurations described in connection with Figures 14C and 14D, although less efficient, may prove useful for performing measurements in highly irradiated primary circuits.

[0129] A study of the thermal and neutron sensitivities of the so-called reference configuration described above, and shown in Figures 12A and 12B, was carried out: lead shielding thickness of 3.5 cm, MgO reflector thickness of 8 cm. This is an optimized configuration compared to the configuration described in relation to [Fig. 5], in particular by the presence of the reflector, the lead shielding, and the optimization of the positioning of the detectors and the source in the cavity.

[0130] Figure 15 shows the thermal sensitivity (curve a - unit: s'.°C') and the neutron sensitivities (unit s'.ppm') for heat transfer fluid temperatures of 277°C (curve b) and 20°C (curve c). The gain in sensitivity compared to Figure 5 can be appreciated. This gain in sensitivity is due to the presence of the reflector, as well as to the position of the detectors and the neutron source, as close as possible to the front end, i.e., as close as possible to the pipe.

[0131] Figure 10A represents a third embodiment, in which the cavity includes auxiliary detectors 19, formed by proportional counters, without boron deposition. The auxiliary detectors are intended to establish an auxiliary detection signal, solely representative of the irradiation γ to which the detectors 14 are subjected. By calibration, an irradiation level can be extracted from the auxiliary detection signal generated by each detector 19. A contribution of the irradiation γ to the detection signal generated by the detectors 14 can then be estimated. This contribution can be subtracted from the detection signal generated by each detector 14, in order to extract the contribution of the detection signal due to the detected neutrons. This assumes knowledge of the response of each detector to the irradiation level y. Other types of auxiliary detectors can be used, for example Geiger-Müller detectors or detectors based on scintillator crystals.

[0132] One advantage of this configuration is that it does not require a dense screen, intended to attenuate gamma radiation.

[0133] Figure [10B] illustrates a variant in which the auxiliary detectors 19 are arranged outside the cavity. This reduces the overall size of the cavity. Thus, the auxiliary detector or detectors 19, as well as the neutron source, can be arranged either inside or outside the cavity.

[0134] The invention can be used to characterize the quantity of boron circulating in the primary circuit of a nuclear reactor, for example a pressurized water reactor.

Claims

Demands

1. Device (1) for measuring a concentration of a neutron absorber in a fluid (3), the device being intended to be disposed facing a pipe (2) of the fluid, the device comprising: - a cavity (10), delimited by a shell (11), the cavity having a front end, intended to be disposed facing the pipe; - a neutron source (13), extending between the shell and the front end; - at least one neutron detector (14), disposed in the cavity, generating a detection signal dependent on the irradiation of the cavity by the neutrons; - a processing unit (20), configured to estimate the concentration of the neutron absorber in the fluid from the detection signal;- the device being characterized in that it comprises a neutron reflector (15), the neutron source being disposed between the front end and the neutron reflector, the neutron reflector being configured to reflect neutrons, emitted by the neutron source, back towards the front end of the cavity.;

2. Device according to claim 1, wherein the neutron source is disposed in the cavity.

3. Device according to claim 1, comprising a front screen (18), disposed between the neutron detector or each detector and the front end, configured to attenuate at least 75% of gamma radiation of energy 1 MeV emitted by the conduit.

4. Device according to claim 3, wherein the front screen is disposed at the front end.

5. Device according to any one of claims 1 to 3, wherein the neutron source (13) is surrounded by the front screen.

6. Device according to any one of the preceding claims, wherein the cavity (10) has an opening at the front end (12).

7. A device according to any one of the preceding claims, wherein the neutron reflector (15) is formed of a material having an albedo greater than 0.7, albedo being a ratio between a flux neutron incident to the material and a neutron flux backscattered by the material.

8. Device according to any one of the preceding claims, wherein the neutron reflector is formed of a material comprising carbon and / or oxygen and / or hydrogen and / or beryllium.

9. Device according to any one of claims 1 to 7, wherein the neutron reflector comprises a material selected from MgO, Graphite, Zr3Si2, MgAl2O4, Y3A150i2, SiC, ZrC, TiC, V-4Ti-4Cr, steel, UO2.

10. Device according to any one of the preceding claims, wherein the neutron reflector extends to a distance from the neutron source of less than 20 cm.

11. Device according to any one of the preceding claims, wherein the cavity extends around a center (0), positioned equidistant between the hull and the front end, the center of the cavity being disposed less than 20 cm from the front end.

12. Device according to any one of the preceding claims, wherein the neutron reflector extends, along the hull, with a thickness greater than 2 cm and less than 20 cm.

13. Device according to any one of the preceding claims, wherein the neutron detector or each neutron detector extends between the neutron source and the shell.

14. Device according to any one of the preceding claims, wherein the reflector extends into or along the hull.

15. A device according to any one of the preceding claims, comprising at least one auxiliary detector (19) configured to generate an auxiliary detection signal dependent on the gamma irradiation level, the processing unit being configured to: - estimate a gamma irradiation level from the auxiliary detection signal; - from the gamma irradiation level, determine a component of the detection signal, resulting from the detector or each detector, relating to the gamma irradiation; - from said component, extract a component of the detection signal, resulting from the detector or each detector, relating to the neutron irradiation.

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