Nuclear power determination method, system, device, medium, and product for a reactor

Abstract

The present disclosure provides a nuclear power determination method, system, device, medium and product of a reactor, the method comprising: obtaining a measured neutron current signal and a subcritical neutron current signal of the reactor; the subcritical neutron current signal is obtained in a subcritical state of the reactor; based on a neutron transport equation constructed based on basic parameters of the reactor, the normalized fission neutron source strength and the average primary fission neutron number of the reactor are calculated respectively; at least based on the subcritical neutron current signal, the normalized fission neutron source strength and the average primary fission neutron number, a power conversion coefficient is established to calculate the measured neutron current signal to obtain the nuclear power; the power conversion coefficient is at least positively correlated with the normalized fission neutron source strength, and negatively correlated with the average primary fission neutron number and the subcritical neutron current signal, which can be used for nuclear power measurement before thermodynamic calibration, i.e. in the low-power stage, and improve the accuracy of nuclear power measurement.

Description

Technical Field

[0001] This disclosure relates to the field of nuclear industry technology, and in particular to a method, system, equipment, medium and product for determining the nuclear power of a reactor. Background Technology

[0002] A nuclear reactor is a complex device that continuously converts nuclear energy into thermal energy through a controlled nuclear fission chain reaction. The nuclear fuel in the reactor core undergoes fission under the induction of thermal neutrons or fast neutrons, releasing enormous energy, which is then carried out by a coolant to generate electricity or provide power.

[0003] In this energy conversion process, nuclear power (i.e., the total amount of heat energy released per unit time within the reactor core) is not only a direct indicator of reactor power level, but also a core physical quantity characterizing the neutron flux level and fission reaction intensity within the reactor core. Accurate measurement of nuclear power is crucial throughout the entire life cycle of the reactor, from startup to full-power operation.

[0004] However, existing nuclear power measurement technologies rely on thermal balance methods for verification under high-power conditions. However, under low-power conditions, it is difficult to calculate the heat carried away by the coolant and the heat leakage, making it impossible to use thermal balance methods to verify nuclear power, thus affecting the accuracy of nuclear power measurement. Summary of the Invention

[0005] The technical problem to be solved by this disclosure is that the nuclear power measurement under low power conditions is inaccurate in the prior art, and provides a method, system, equipment, medium and product for determining the nuclear power of a reactor.

[0006] This disclosure solves the above-mentioned technical problems through the following technical solution:

[0007] In a first aspect, this disclosure provides a method for determining the nuclear power of a reactor, the method comprising:

[0008] The measured neutron current signal and subcritical neutron current signal of the reactor were acquired; the subcritical neutron current signal was acquired under the subcritical state of the reactor.

[0009] The neutron transport equations constructed based on the reactor's fundamental parameters were used to calculate the normalized fission neutron source strength and the average number of neutrons per fission.

[0010] A power conversion coefficient is established based at least on the subcritical neutron current signal, the normalized fission neutron source intensity, and the average primary fission neutron number to calculate the nuclear power from the measured neutron current signal. The power conversion coefficient is at least positively correlated with the normalized fission neutron source intensity and negatively correlated with the average primary fission neutron number and the subcritical neutron current signal.

[0011] Optionally, when the reactor is in a subcritical state, the measured neutron current signal is obtained by converting the pulse signal collected by the pulse detector, or by obtaining the current signal collected by the current detector.

[0012] Optionally, when the reactor is in a critical state, the neutron current signal is obtained based on a current detector.

[0013] Optionally, the power conversion coefficient also includes the primary fission energy and the external neutron source intensity; the power conversion coefficient is positively correlated with both the primary fission energy and the external neutron source intensity.

[0014] Power conversion factors include:

[0015] ;

[0016] Where Z is the power conversion coefficient and Q is the energy of a single fission. S represents the normalized fission neutron source intensity, and S represents the external neutron source intensity. The average number of neutrons per fission cycle. This is the subcritical neutron current signal.

[0017] Optionally, the basic parameters include at least the number of neutrons in a single micro-fission, the macro-fission cross section, and the spatial volume parameters;

[0018] The neutron transport equations include a fixed-source computational model, which includes:

[0019] ;

[0020] Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. Strong external neutron source;

[0021] The normalized fission neutron source intensity is calculated based on the neutron flux density distribution obtained from the fixed-source calculation model. The calculation formula includes:

[0022] ;

[0023] in, To normalize the fission neutron source strength, The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

[0024] Optionally, the basic parameters include at least the effective multiplication factor, the number of microscopic single fission neutrons, the macroscopic fission cross section, and the spatial volume parameters;

[0025] The neutron transport equations include a critical computational model, which includes:

[0026] ;

[0027] Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. The effective proliferation factor;

[0028] The average number of neutrons in a single fission is calculated based on the neutron flux density distribution obtained from the critical calculation model. The calculation formula includes:

[0029] ;

[0030] in, The average number of neutrons per fission cycle. The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

[0031] Optionally, the measured neutron current signal and subcritical neutron current signal of the reactor are acquired, including:

[0032] The subcritical neutron current signal of the reactor was acquired multiple times during the subcritical state of the reactor.

[0033] The neutron transport equations constructed based on the reactor's fundamental parameters were used to calculate the normalized fission neutron source strength and the average number of neutrons per primary fission, including:

[0034] Under the subcritical state of the reactor, the normalized fission neutron source intensity of the reactor was calculated multiple times based on the neutron transport equation constructed from the basic parameters of the reactor.

[0035] The method also includes:

[0036] Calculate the ratio of each critical neutron current signal to the normalized fission neutron source intensity, and calculate the average value based on all ratios;

[0037] The power conversion coefficient is established based at least on the subcritical neutron current signal, the normalized fission neutron source intensity, and the average primary fission neutron number. The nuclear power is then calculated from the measured neutron current signal, including:

[0038] A power conversion coefficient is established based on the average value and the average number of neutrons in a single fission to calculate the nuclear power from the measured neutron current signal; the nuclear power is positively correlated with the average value.

[0039] Secondly, this disclosure provides a reactor nuclear power determination system, the system comprising:

[0040] The first acquisition module is used to acquire the measured neutron current signal and subcritical neutron current signal of the reactor; the subcritical neutron current signal is acquired under the subcritical state of the reactor.

[0041] The first calculation module is used to construct neutron transport equations based on the reactor's basic parameters, and calculate the normalized fission neutron source strength and the average number of neutrons per fission in the reactor, respectively.

[0042] The second calculation module is used to establish a power conversion coefficient based at least on the subcritical neutron current signal, the normalized fission neutron source strength, and the average primary fission neutron number, in order to calculate the nuclear power from the measured neutron current signal; the power conversion coefficient is at least positively correlated with the normalized fission neutron source strength and negatively correlated with the average primary fission neutron number and the subcritical neutron current signal.

[0043] Thirdly, this disclosure provides an electronic device including a memory, a processor, and a computer program stored in the memory and for running on the processor, wherein the processor executes the computer program to implement a method for determining the nuclear power of a reactor as described in any of the first aspects.

[0044] Fourthly, this disclosure provides a computer-readable storage medium having a computer program stored thereon, which, when executed by a processor, implements a method for determining the nuclear power of a reactor as described in any of the first aspects.

[0045] Fifthly, this disclosure provides a computer program product, including a computer program that, when executed by a processor, implements a method for determining the nuclear power of a reactor as described in any of the first aspects.

[0046] Based on common knowledge in the field, the above-mentioned preferred conditions can be combined arbitrarily to obtain various preferred embodiments of this disclosure.

[0047] The positive and progressive effects of this disclosure are as follows: due to the low power or even subcritical loading of the reactor, there are very few spontaneous fission neutrons in the reactor. In this embodiment, under the subcritical state of the nuclear reactor, an external neutron source of known intensity is introduced. Under this specific physical state, all the neutron flux in the reactor is excited by this external neutron source of known intensity through a finite number of multiplications in the subcritical system. At this time, the subcritical neutron current can be measured, providing a calibration method for determining nuclear power that does not depend on any thermodynamic calibration. Meanwhile, nuclear power cannot be directly derived from subcritical neutron current alone. This disclosure employs the neutron transport equation to solve for the normalized fission neutron source strength and the average primary fission neutron number. The normalized fission neutron source strength reflects the spatial proportionality between the probe neutron current signal and the overall core fission rate, quantifying whether actual nuclear fission occurred in the core under the drive of the external neutron source. The average primary fission neutron number reflects the correspondence between neutron production and the microscopic physical properties of nuclides. By combining the external neutron source with the actual core fission through these two parameters, the dependence on thermodynamic calibration is reduced, and the accuracy of nuclear power measurement is improved. Finally, a power conversion coefficient is constructed based on the above parameters, enabling accurate calculation of nuclear power. In summary, by utilizing the multiplication effect of the external neutron source under subcritical conditions and high-precision numerical calculations, a power conversion coefficient is constructed, thereby endowing the external detector with the ability to perform absolute power measurement in the very early stages of reactor startup. Attached Figure Description

[0048] Figure 1 A flowchart illustrating a method for determining the nuclear power of a reactor, provided as an exemplary embodiment of this disclosure;

[0049] Figure 2 A structural diagram of a nuclear power determination system for a reactor provided as an exemplary embodiment of this disclosure;

[0050] Figure 3 This is a structural diagram of an electronic device provided as an exemplary embodiment of the present disclosure. Detailed Implementation

[0051] The present disclosure is further illustrated below by way of embodiments, but the present disclosure is not limited to the scope of the embodiments.

[0052] The prefixes such as "first" and "second" used in this disclosure are merely for distinguishing different descriptive objects and do not limit the position, order, priority, quantity, or content of the described objects. The use of ordinal numbers and other prefixes used to distinguish descriptive objects in this disclosure does not constitute a limitation on the described objects. The description of the described objects is given in the claims or the context of the embodiments, and should not be construed as an unnecessary limitation. Furthermore, in the description of this embodiment, unless otherwise stated, "multiple" means two or more.

[0053] During their research on reactor nuclear power, the inventors discovered the following method for determining nuclear power:

[0054] (1) Thermal balance method.

[0055] The core thermal power is calculated using the heat balance equation by measuring the flow rate, inlet and outlet temperature difference, and physical properties of the coolant in the reactor loop. This method is highly accurate, but it heavily relies on the coolant generating a stable and measurable temperature rise. At low power levels in the reactor (such as below 1% of rated power), the temperature difference signal is extremely weak, and the measurement uncertainty increases significantly. Therefore, it cannot be used to determine the power during the initial stage of reactor startup.

[0056] (2) Off-pile neutron flux measurement method.

[0057] Detectors such as ionization chambers and proportional counters are installed outside the pressure vessel to continuously monitor neutron flux based on the proportional relationship between neutron flux and nuclear power. This method offers fast response and allows for online monitoring. However, there is actually no direct physical proportional relationship between the neutron detector output signal and nuclear power; calibration must rely on other absolute measurement methods. During the reactor's initial criticality or low-power physics tests, due to the lack of a calibration benchmark, neutron detector readings can only reflect relative changes and cannot provide accurate absolute power values.

[0058] (3) Activation method.

[0059] A known mass of fissile material (such as gold foil or dysprosium foil) is placed in the reactor core and irradiated for a certain period of time. The absolute fission rate is then calculated by measuring the activity of the generated radionuclides, thus obtaining the nuclear power. This method is highly accurate, but it is an offline and destructive measurement, making real-time continuous monitoring impossible. Furthermore, it is complex and time-consuming, making it unsuitable as a routine power monitoring method.

[0060] (4) Neutron noise analysis method.

[0061] By utilizing the inherent statistical fluctuations (neutron noise) of neutron flux, dynamic parameters of the reactor are extracted and nuclear power is estimated by analyzing the power spectral density of the neutron detector signal. This method can operate at low power (even subcritical) conditions and does not depend on the absolute efficiency of the detector, but it requires complex signal processing and theoretical models, and is highly susceptible to external vibration and noise interference, making its engineering implementation difficult.

[0062] (5) Neutron source introduction method.

[0063] In the subcritical state, an external neutron source of known intensity is introduced into the reactor core, and the nuclear power is inferred by measuring changes in the neutron detector signal (such as the linear growth slope or multiplication factor). However, this method relies on corrections for the source value and reactivity changes during the source introduction process, and the theoretical and experimental procedures are complex.

[0064] As can be seen from the above methods, these methods for determining nuclear power have significant shortcomings in the low-power region (especially under subcritical conditions): the thermal equilibrium method fails, the off-pile neutron flux measurement method lacks an absolute reference, the activation method is complex and not real-time, and the neutron noise analysis method and the neutron source introduction method have complex theoretical models and are difficult to implement in engineering. Therefore, there is an urgent need for a method that can absolutely calibrate the nuclear measurement system under low-power conditions to accurately determine the nuclear power.

[0065] Based on this Figure 1 A flowchart illustrating an exemplary embodiment of this disclosure provides a method for determining the nuclear power of a reactor, the method comprising:

[0066] S101. Obtain the measured neutron current signal and subcritical neutron current signal of the reactor.

[0067] The subcritical neutron current signal was acquired during the subcritical state of the reactor. The reactor state includes, but is not limited to, the critical state and the subcritical state, based on the effective multiplication factor. To characterize, theoretically speaking, This is a critical state. This is a subcritical state. This embodiment considers the selection of the subcritical state as a supercritical state. The closer to 1, the better; the preferred value is... The current corresponds to the subcritical state, because in this state, the current signal's response to the external neutron source is negligible relative to the fission source, and the current signal's response to the internal fission source also tends to be stable.

[0068] Specifically, in the subcritical state of the reactor, the measured neutron current signal and the subcritical neutron current signal can be obtained based on either a pulse detector or a current detector. Specifically, if the neutron current signal is obtained by converting the pulse signal from the pulse detector, the general method is to first acquire the pulse signal from the pulse detector in the subcritical state, and then convert the pulse signal into the corresponding neutron current signal using the proportional relationship between the pulse detector and the current detector during the connection process. If the neutron current signal is acquired based on the current detector, the current signal detected by the current detector can be directly output as the neutron current signal.

[0069] Furthermore, under the critical state of the reactor, the measured neutron current signal can be directly obtained based on the current signal collected by the current detector.

[0070] S102. Based on the reactor's fundamental parameters, the neutron transport equations were constructed, and the normalized fission neutron source strength and the average number of neutrons per fission were calculated.

[0071] The neutron transport equation can be constructed based on any of the deterministic methods, Monte Carlo methods, etc. For details, please refer to any of the existing technical means. In this embodiment, the construction of the neutron transport equation will not be the focus of the discussion.

[0072] The fundamental parameters of a reactor include at least one or more of the following: microscopic single-fission neutron number, macroscopic fission cross section, spatial volume parameters, primary fission energy, and external neutron source strength. These fundamental parameters, such as the microscopic single-fission neutron number and microscopic fission cross section, can be obtained from a nuclear database. Spatial volume parameters can be determined based on the reactor's engineering design. The macroscopic fission cross section is calculated based on the microscopic fission cross section. The primary fission energy is related to the fuel; for example, U-235 fuel has a value of approximately 200 MeV. The external neutron source strength can be directly measured from an external neutron source. These parameters can be automatically obtained by a program running neutron transport equations, and will not be elaborated upon in this embodiment. Based on these fundamental parameters, the calculation process of the normalized fission neutron source strength and the average primary fission neutron number in step S102 of this embodiment is illustrated. It should be noted that the calculation process provided in this embodiment is only for reference, and those skilled in the art should know that other methods can be used to calculate the normalized fission neutron source strength and the average primary fission neutron number.

[0073] Specifically, regarding the normalized fission neutron source intensity, the neutron transport equation can be solved using a fixed-source computational model. The normalized fission neutron source intensity characterizes the total steady-state fission intensity achieved by the reactor in a subcritical state after being excited by a specific external neutron source (such as an Am-Be source) and undergoing a finite number of chain reaction amplifications. It depends on the location and intensity of the external source. Therefore, a fixed-source computational model (i.e., a non-homogeneous equation with an external source term) is needed to realistically simulate the physical process of external neutron injection into the reactor core. Based on this, the fixed-source computational model includes:

[0074] ;

[0075] Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. The external neutron source strength is given. The neutron migration and loss operator and the neutron fission generation operator are calculated based on the above-mentioned basic parameters, and the specific calculation methods can be found in the prior art. The strength of the external neutron source is specifically obtained through the collection of external neutron sources.

[0076] After obtaining the neutron flux density distribution, the normalized fission neutron source intensity is calculated based on the neutron flux density distribution obtained from the fixed-source calculation model. The calculation formula includes:

[0077] ;

[0078] in, To normalize the fission neutron source strength The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

[0079] Specifically, regarding the average number of neutrons per fission cycle, the neutron transport equation can be solved using a critical calculation model. The average number of neutrons per fission cycle represents the statistical average of neutrons released during a single macroscopic fission under a specific reactor geometry and material composition. It is an intrinsic property of the reactor and is unrelated to the presence of external sources. Therefore, to obtain the true intrinsic properties of the reactor core, it is necessary to eliminate the interference of external sources and employ a critical calculation model (i.e., a homogeneous eigenvalue equation without external sources). Based on this, the critical calculation model includes:

[0080] ;

[0081] Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. For effective proliferation factor, It is calculated based on the ratio of the total number of neutrons produced by the next generation of fission to the total number of neutrons produced by the previous generation of fission.

[0082] After obtaining the neutron flux density distribution, the average number of neutrons per fission cycle is calculated based on the neutron flux density distribution obtained from the critical calculation model. The calculation formula includes:

[0083] ;

[0084] in, The average number of neutrons per fission cycle. The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

[0085] S103. At least based on the subcritical neutron current signal, the normalized fission neutron source strength, and the average number of primary fission neutrons, establish a power conversion coefficient to calculate the nuclear power from the measured neutron current signal.

[0086] Among them, the power conversion coefficient is at least positively correlated with the normalized fission neutron source strength, and negatively correlated with the average primary fission neutron number and the subcritical neutron current signal.

[0087] In addition to establishing the power conversion coefficient from the parameters mentioned above, the power conversion coefficient also includes correction parameters, which are constant values ​​and can generally be obtained empirically. Alternatively, the power conversion coefficient may also include the primary fission energy and the external neutron source intensity, and both the primary fission energy and the external neutron source intensity are positively correlated with the power conversion coefficient. As just one example, the power conversion coefficient includes:

[0088] ;

[0089] Where Z is the power conversion coefficient and Q is the energy of a single fission. S represents the normalized fission neutron source intensity, and S represents the external neutron source intensity. The average number of neutrons per fission cycle. This is the subcritical neutron current signal.

[0090] After obtaining the power conversion coefficient, the formula for calculating the nuclear power in step S103 is shown below:

[0091] ;

[0092] Where P is nuclear power, I is measured neutron current, and Z is power conversion factor.

[0093] In one embodiment, to improve the measurement accuracy of nuclear power, it can be performed under subcritical conditions (preferably at different...). (Characterizing the situation), multiple subcritical neutron current signals and multiple normalized fission neutron source intensities are obtained, and the average value of their ratios is calculated. Then, the power conversion coefficient is constructed based on the average value.

[0094] Specifically, step S101 includes: acquiring the subcritical neutron current signal of the reactor multiple times during the subcritical state of the reactor. Step S102 includes: calculating the normalized fission neutron source intensity of the reactor multiple times during the subcritical state of the reactor based on the neutron transport equation constructed from the reactor's fundamental parameters. The method also includes: calculating the ratio of each subcritical neutron current signal to the normalized fission neutron source intensity, and calculating the average value based on all ratios. Step S103 includes: establishing a power conversion coefficient based on the average value and the average number of primary fission neutrons to calculate the nuclear power from the measured neutron current signal; the nuclear power is positively correlated with the average value.

[0095] In one embodiment, the method further includes: modifying the neutron transport model through methods such as critical extrapolation, subcritical reactivity measurement, and chemical analysis.

[0096] The method for determining the nuclear power of a reactor provided in this embodiment will be explained in detail below through a specific example:

[0097] Taking a molten salt reactor as an example, the fissile nuclide U-235 is used in the fuel, the carrier is FLiBe molten salt, and graphite is used as the moderator. The rated full power is 200 MWth. The U-235 source is enriched uranium (U-235 mass percentage is 19.75 wt%). The external neutron source is an Am-Be source with a source intensity of 4E7n / s.

[0098] Measuring the reactor at The subcritical neutron current collected by the detector at a neutron density of 0.99786 was 1.5E-11 A. Next, using the Monte Carlo program and a fixed-source calculation model, the subcritical neutron current was calculated... At a neutron concentration of 0.99786, the normalized fission neutron source intensity of the reactor is 246.27 n / s. Using the criticality calculation model, the average number of neutrons per fission cycle in the reactor is calculated to be 2.43. Here, the Monte Carlo calculation model can be corrected by extrapolating the critical uranium concentration to ensure that the calculation model corresponds to the actual reactor conditions.

[0099] Based on the formula in the above embodiments, the power conversion coefficient between the reactor nuclear power and the detector current is calculated to be 8.65E9W / A. This power coefficient can be used to obtain the reactor nuclear power under any neutron detector current. For example, when the measured current signal is 2.31E-2A, the reactor nuclear power is 200MW, meaning the reactor is operating at its rated full power.

[0100] In this embodiment, due to the extremely low number of spontaneous fission neutrons in the reactor under low power or even subcritical conditions, a known external neutron source is introduced into the subcritical state of the nuclear reactor. Under this specific physical state, all neutron flux within the reactor is excited by this known-intensity external neutron source through a finite number of multiplications within the subcritical system. At this point, the subcritical neutron current can be measured, providing a method for determining nuclear power that is independent of any thermodynamic calibration. However, nuclear power cannot be directly derived from the subcritical neutron current alone. This embodiment uses the neutron transport equation to solve for the normalized fission neutron source strength and the average number of primary fission neutrons. The normalized fission neutron source strength reflects the spatial proportionality between the probe neutron current signal and the overall fission rate of the entire reactor core, quantifying whether actual nuclear fission has occurred in the reactor core under the drive of the external neutron source. The average number of primary fission neutrons reflects the correspondence between neutron production and the microscopic physical properties of nuclides. By combining these two parameters, the external neutron source and the actual core fission are integrated, reducing the dependence on thermodynamic calibration. Finally, based on the above parameters, a power conversion coefficient is constructed, enabling accurate calculation of nuclear power. In summary, by utilizing the multiplication effect of the external neutron source under subcritical conditions and high-precision numerical calculations, a power conversion coefficient is constructed, thereby endowing the off-chip detector with the ability to perform absolute power measurements in the very early stages of reactor startup.

[0101] Corresponding to the aforementioned embodiments of the reactor nuclear power determination method, this disclosure also provides embodiments of a reactor nuclear power determination system.

[0102] Figure 2 A schematic diagram of a nuclear power determination system for a reactor, provided as an exemplary embodiment of this disclosure, is shown. The system includes:

[0103] The first acquisition module 21 is used to acquire the measured neutron current signal and subcritical neutron current signal of the reactor. The subcritical neutron current signal is acquired under the subcritical state of the reactor.

[0104] The first calculation module 22 is used to calculate the normalized fission neutron source strength and the average number of fission neutrons per cycle of the reactor based on the neutron transport equation constructed from the reactor's basic parameters.

[0105] The second calculation module 23 is used to establish a power conversion coefficient based at least on the subcritical neutron current signal, the normalized fission neutron source intensity, and the average primary fission neutron number, in order to calculate the nuclear power from the measured neutron current signal. The power conversion coefficient is at least positively correlated with the normalized fission neutron source intensity and negatively correlated with the average primary fission neutron number and the subcritical neutron current signal.

[0106] In one embodiment, when the reactor is in a critical or supercritical state, the measured neutron current signal is obtained based on the current signal collected by the neutron detector.

[0107] In one embodiment, when the reactor is in a subcritical state, the measured neutron current signal is obtained by converting the pulse signal collected by the pulse detector.

[0108] In one embodiment, the power conversion coefficient also includes the primary fission energy and the external neutron source intensity. The power conversion coefficient is positively correlated with both the primary fission energy and the external neutron source intensity.

[0109] Power conversion factors include:

[0110] .

[0111] Where Z is the power conversion coefficient and Q is the energy of a single fission. S represents the normalized fission neutron source intensity, and S represents the external neutron source intensity. The average number of neutrons per fission cycle. This is the subcritical neutron current signal.

[0112] In one embodiment, the basic parameters include at least the number of microscopic single fission neutrons, the macroscopic fission cross section, and the spatial volume parameters.

[0113] The neutron transport equations include a fixed-source computational model, which includes:

[0114] .

[0115] Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. It is a strong external neutron source.

[0116] The normalized fission neutron source intensity is calculated based on the neutron flux density distribution obtained from the fixed-source calculation model. The calculation formula includes:

[0117] .

[0118] in, To normalize the fission neutron source strength The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

[0119] In one embodiment, the basic parameters include at least the effective multiplication factor, the number of microscopic single fission neutrons, the macroscopic fission cross section, and the spatial volume parameter.

[0120] The neutron transport equations include a critical computational model, which includes:

[0121] .

[0122] Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. This is the effective proliferation factor.

[0123] The average number of neutrons in a single fission is calculated based on the neutron flux density distribution obtained from the critical calculation model. The calculation formula includes:

[0124] .

[0125] in, The average number of neutrons per fission cycle. The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

[0126] In one embodiment, the first acquisition module 21 is further configured to acquire the subcritical neutron current signal of the reactor multiple times during the subcritical state of the reactor.

[0127] The first calculation module 22 is also used to calculate the normalized fission neutron source intensity of the reactor multiple times based on the neutron transport equation constructed from the basic parameters of the reactor under the subcritical state of the reactor.

[0128] The system also includes a third calculation module, which calculates the ratio of each critical neutron current signal to the normalized fission neutron source intensity, and calculates the average value based on all ratios.

[0129] The second calculation module 23 is also used to establish a power conversion coefficient based on the average value and the average number of neutrons in a single fission, and to calculate the nuclear power from the measured neutron current signal. The nuclear power is positively correlated with the average value.

[0130] For the system embodiments, since they basically correspond to the method embodiments, the relevant parts can be referred to in the description of the method embodiments. The system embodiments described above are merely illustrative, wherein the units described as separate components may or may not be physically separate, and the components as units may or may not be physical units, that is, they may be located in one place or distributed across multiple network units. Some or all of the modules can be selected to achieve the purpose of this disclosure solution according to actual needs.

[0131] Figure 3This is a schematic diagram of the structure of an electronic device according to an example embodiment of the present disclosure. The electronic device includes a memory, a processor, and a computer program stored in the memory and used to run on the processor. When the processor executes the computer program, it implements the nuclear power determination method of the reactor in any of the above embodiments. Figure 3 The electronic device 30 shown is merely an example and should not impose any limitation on the functionality and scope of use of the embodiments disclosed herein.

[0132] like Figure 3 As shown, the electronic device 30 can be manifested as a general-purpose computing device, such as a server device. The components of the electronic device 30 may include, but are not limited to: at least one processor 31, at least one memory 32, and a bus 33 connecting different system components (including memory 32 and processor 31).

[0133] Bus 33 includes a data bus, an address bus, and a control bus.

[0134] The memory 32 may include volatile memory, such as random access memory (RAM) 321 and / or cache memory 322, and may further include read-only memory (ROM) 323.

[0135] The memory 32 may also include a program tool 325 (or utility) having a set (at least one) program module 324, such program module 324 including but not limited to: an operating system, one or more application programs, other program modules and program data, each or some combination of these examples may include an implementation of a network environment.

[0136] The processor 31 executes various functional applications and data processing by running computer programs stored in the memory 32, such as the reactor nuclear power determination method provided in any of the above embodiments.

[0137] Electronic device 30 can also communicate with one or more external devices 34 (e.g., keyboard, pointing device, etc.). This communication can be performed via input / output (I / O) interface 35. Furthermore, electronic device 30 can also communicate with one or more networks (e.g., local area network (LAN), wide area network (WAN), and / or public networks, such as the Internet) via network adapter 36. Figure 3 As shown, network adapter 36 communicates with other modules of electronic device 30 via bus 33. It should be understood that, although... Figure 3 Not shown, other hardware and / or software modules may be used in conjunction with electronic device 30, including but not limited to: microcode, device drivers, redundant processors, external disk drive arrays, RAID (disk array) systems, tape drives, and data backup storage systems.

[0138] It should be noted that although several units / modules or sub-units / modules of the electronic device have been mentioned in the detailed description above, this division is merely exemplary and not mandatory. In fact, according to embodiments of this disclosure, the features and functions of two or more units / modules described above can be embodied in one unit / module. Conversely, the features and functions of one unit / module described above can be further divided and embodied by multiple units / modules.

[0139] This disclosure also provides a computer-readable storage medium storing a computer program thereon, which, when executed by a processor, implements the reactor nuclear power determination method provided in any of the above embodiments.

[0140] The readable storage medium may be more specifically adopted, including but not limited to: portable disk, hard disk, random access memory, read-only memory, erasable programmable read-only memory, optical storage device, magnetic storage device, or any suitable combination thereof.

[0141] This disclosure also provides a computer program product, including a computer program that, when executed by a processor, implements the reactor nuclear power determination method described in any of the preceding embodiments.

[0142] The program code for executing the computer program product of this disclosure can be written in any combination of one or more programming languages, and the program code can be executed entirely on a user device, partially on a user device, as a stand-alone software package, partially on a user device and partially on a remote device, or entirely on a remote device.

[0143] While specific embodiments of this disclosure have been described above, those skilled in the art should understand that these are merely illustrative examples, and the scope of protection of this disclosure is defined by the appended claims. Those skilled in the art can make various changes or modifications to these embodiments without departing from the principles and essence of this disclosure, but all such changes and modifications fall within the scope of protection of this disclosure.

Claims

1. A method for determining the nuclear power of a reactor, characterized in that, The method includes: The measured neutron current signal and subcritical neutron current signal of the reactor are acquired; the subcritical neutron current signal is acquired under the subcritical state of the reactor. The neutron transport equations constructed based on the reactor's fundamental parameters were used to calculate the normalized fission neutron source strength and the average number of neutrons per fission cycle of the reactor. A power conversion coefficient is established based at least on the subcritical neutron current signal, the normalized fission neutron source strength, and the average primary fission neutron number to calculate the nuclear power from the measured neutron current signal; the power conversion coefficient is at least positively correlated with the normalized fission neutron source strength and negatively correlated with the average primary fission neutron number and the subcritical neutron current signal.

2. The method for determining nuclear power as described in claim 1, characterized in that, When the reactor is in a subcritical state, the measured neutron current signal is obtained by converting the pulse signal collected by the pulse detector. Alternatively, when the reactor is in a subcritical state, the measured neutron current signal is obtained based on a current detector. Alternatively, when the reactor is in a critical state, the measured neutron current signal is obtained based on a current detector.

3. The method for determining nuclear power as described in claim 1 or 2, characterized in that, The power conversion coefficient also includes the primary fission energy and the external neutron source intensity; the power conversion coefficient is positively correlated with both the primary fission energy and the external neutron source intensity. The power conversion coefficient includes: ； Where Z is the power conversion coefficient and Q is the energy of a single fission. S represents the normalized fission neutron source intensity, and S represents the external neutron source intensity. The average number of neutrons per fission cycle. This is the subcritical neutron current signal.

4. The method for determining nuclear power as described in claim 1, characterized in that, The basic parameters include at least the number of neutrons in a single micro-fission, the macro-fission cross section, and the spatial volume parameters. The neutron transport equation includes a fixed-source computational model, which includes: ； Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. Strong external neutron source; The normalized fission neutron source strength is calculated based on the neutron flux density distribution obtained from the fixed source calculation model, and the calculation formula includes: ； in, The normalized fission neutron source strength is... The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

5. The method for determining nuclear power as described in claim 1, characterized in that, The basic parameters include at least the effective multiplication factor, the number of neutrons in a single micro-fission, the macro-fission cross section, and the spatial volume parameters. The neutron transport equation includes a critical calculation model, which includes: ； Where L is the neutron migration and loss operator, and M is the neutron fission production operator. This represents the flux density distribution of neutrons at space r and energy E. The effective proliferation factor; The average number of neutrons per primary fission is calculated based on the neutron flux density distribution obtained from the critical calculation model, and the calculation formula includes: ； in, The average number of neutrons per fission cycle, The number of neutrons in a single microscopic fission at energy E. Let be the macroscopic fission cross section of a neutron at space r and energy E, and V be the spatial volume parameter.

6. The method for determining nuclear power as described in claim 1, characterized in that, The acquisition of the measured neutron current signal and subcritical neutron current signal of the reactor includes: In the subcritical state of the reactor, the subcritical neutron current signal of the reactor is acquired multiple times. The neutron transport equations constructed based on the reactor's fundamental parameters are used to calculate the normalized fission neutron source strength and the average number of neutrons per fission event of the reactor, including: Under the subcritical state of the reactor, the normalized fission neutron source intensity of the reactor was calculated multiple times based on the neutron transport equation constructed from the basic parameters of the reactor. The method further includes: Calculate the ratio of the subcritical neutron current signal to the normalized fission neutron source intensity for each time, and calculate the average value based on all the ratios; The process of establishing a power conversion coefficient based at least on the subcritical neutron current signal, the normalized fission neutron source strength, and the average primary fission neutron number to calculate the nuclear power from the measured neutron current signal includes: A power conversion coefficient is established based on the average value and the average number of neutrons in a single fission. The nuclear power is calculated from the measured neutron current signal. The nuclear power is positively correlated with the average value.

7. A reactor nuclear power determination system, characterized in that, The system includes: The first acquisition module is used to acquire the measured neutron current signal and subcritical neutron current signal of the reactor; the subcritical neutron current signal is acquired under the subcritical state of the reactor. The first calculation module is used to calculate the normalized fission neutron source strength and the average number of neutrons per fission cycle of the reactor based on the neutron transport equation constructed from the basic parameters of the reactor. The second calculation module is used to establish a power conversion coefficient based at least on the subcritical neutron current signal, the normalized fission neutron source strength, and the average primary fission neutron number, in order to calculate the nuclear power from the measured neutron current signal; the power conversion coefficient is at least positively correlated with the normalized fission neutron source strength and negatively correlated with the average primary fission neutron number and the subcritical neutron current signal.

8. An electronic device comprising a memory, a processor, and a computer program stored in the memory and for running on the processor, characterized in that, When the processor executes the computer program, it implements the method for determining the nuclear power of the reactor as described in any one of claims 1-6.

9. A computer-readable storage medium having a computer program stored thereon, characterized in that, When the computer program is executed by the processor, it implements the method for determining the nuclear power of the reactor as described in any one of claims 1-6.

10. A computer program product, comprising a computer program, characterized in that, When the computer program is executed by the processor, it implements the method for determining the nuclear power of the reactor as described in any one of claims 1-6.

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