CN114169164B - Method and device for determining core power of critical device - Google Patents
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- QSGNKXDSTRDWKA-UHFFFAOYSA-N zirconium dihydride Chemical compound [ZrH2] QSGNKXDSTRDWKA-UHFFFAOYSA-N 0.000 description 1
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Abstract
The embodiment of the invention discloses a method and a device for determining core power of a critical device, wherein the method comprises the following steps: simulating the critical device by using a Monte Carlo method to obtain a simulation device of the critical device, and representing the relative core power of the simulation device; characterizing a relative thermal neutron flux density at a predetermined location using a monte carlo method; characterizing a relationship between core power of the critical device and thermal neutron flux density at the predetermined location based on the relative core power of the simulation device and the relative thermal neutron flux density at the predetermined location; simulating the operation process of the simulation device by using a Monte Carlo method to calculate the relative thermal neutron flux density at a preset position and the relative core power of the simulation device, and calculating a relation coefficient according to the relation between the core power of the critical device and the thermal neutron flux density at the preset position; starting a critical device; and measuring the thermal neutron flux density at a preset position, and determining the core power of the critical device according to the relation coefficient.
Description
Technical Field
The invention relates to the technical field of nuclear reactors, in particular to a method and a device for determining core power of a critical device.
Background
The critical device is a physical experimental device used for carrying out critical experimental measurement on various arrangement modes and compositions of nuclear fuel and other materials forming a reactor core in a design stage, determining critical characteristics of the critical device and providing basis for verification theory calculation. The critical apparatus can maintain a controllable chain reaction at a low power level and provide conditions for studying core placement and composition.
When the critical device operates, an experimenter needs to know the power of the critical device in real time, so that the critical device can be ensured to operate safely and stably under different experimental conditions.
Disclosure of Invention
The present invention has been made in view of the above problems, and it is an object of the present invention to provide a method and apparatus for determining core power of a critical apparatus that overcomes or at least partially solves the above problems.
A first aspect of an embodiment of the invention provides a method of determining core power of a critical device, the method comprising: simulating a critical device by using a Monte Carlo method to obtain a simulation device of the critical device, and representing the relative reactor core power of the simulation device; characterizing the relative thermal neutron flux density of a preset position by using a Monte Carlo method, wherein the preset position is arranged outside a reactor core of the critical device, and the relative thermal neutron flux density of the preset position is the relative thermal neutron flux density of neutrons generated by the reactor core; characterizing a relationship between core power of the critical device and thermal neutron flux density at the predetermined location based on the relative core power of the simulation device and the relative thermal neutron flux density at the predetermined location; simulating the operation process of the simulation device by using a Monte Carlo method to calculate the relative thermal neutron flux density of the preset position and the relative core power of the simulation device, and calculating a relation coefficient according to the relation between the core power of the simulation device and the thermal neutron flux density of the preset position; starting the critical device; and measuring the thermal neutron flux density of the preset position, and determining the core power of the critical device according to the relation coefficient.
A second aspect of an embodiment of the present invention provides an apparatus for determining core power of a critical apparatus, the apparatus comprising: the simulation part is used for simulating the critical device to obtain a simulation device of the critical device, representing the relative reactor core power of the simulation device and obtaining the simulation device of the critical device; a characterization part, which is used for characterizing the relative thermal neutron flux density of a preset position, and is also used for characterizing the relation between the core power of the critical device and the thermal neutron flux density of the preset position according to the relative core power of the simulation device and the relative thermal neutron flux density of the preset position; a measuring section for measuring a thermal neutron flux density at the predetermined position; and a calculating unit configured to calculate a relationship coefficient from a relationship between core power of the critical device and thermal neutron flux density at the predetermined position.
A third aspect of the invention provides a computer readable storage medium comprising instructions which, when run on a computer, cause the computer to perform a method of determining core power of a critical device as provided by the first aspect of an embodiment of the invention.
Drawings
Other objects and advantages of the present invention will become apparent from the following description of the invention with reference to the accompanying drawings, which provide a thorough understanding of the present invention.
FIG. 1 is a schematic flow chart of a method of determining core power of a critical device according to one embodiment of the invention;
FIG. 2 is a schematic flow chart of step 30 of the method of determining core power of the critical device shown in FIG. 1;
FIG. 3 is a schematic flow chart of step 31 of the method of determining core power of the critical device shown in FIG. 2;
FIG. 4 is a schematic structural view of an apparatus for determining core power of a critical apparatus according to one embodiment of the present invention;
fig. 5 is a schematic structural view of a measuring section according to an embodiment of the present invention.
It should be noted that the figures are not drawn to scale and that elements of similar structures or functions are generally represented by like reference numerals throughout the figures for illustrative purposes. It should also be noted that the drawings are only for the purpose of describing the preferred embodiments and are not intended to limit the invention itself. The drawings do not illustrate every aspect of the described embodiments and do not limit the scope of the invention.
In the figure, 100 is an analog part, 200 is a characterization part, 300 is a measurement part, 301 is a positive voltage cavity, 302 is a negative voltage cavity, 303 is a boron-coated electrode, 304 is a collecting electrode, and 400 is a calculation part.
Detailed Description
In order to make the objects, technical solutions and advantages of the present invention more apparent, the technical solutions of the present invention will be clearly and completely described below with reference to the accompanying drawings of the embodiments of the present invention. It will be apparent that the described embodiments are one embodiment, but not all embodiments, of the present invention. All other embodiments, which can be made by a person skilled in the art without creative efforts, based on the described embodiments of the present invention fall within the protection scope of the present invention.
Unless defined otherwise, technical or scientific terms used herein should be given the ordinary meaning as understood by one of ordinary skill in the art to which this invention belongs.
The critical device is a physical experimental device used for carrying out critical experimental measurement on various arrangement modes and compositions of nuclear fuel and other materials forming a reactor core in a design stage, determining critical characteristics of the critical device and providing basis for verification theory calculation. The critical device has fissionable materials therein, can maintain a controllable chain reaction at low power levels, and provides conditions for studying core placement and composition.
The core power of the critical device can be obtained from the fission rate of the core, i.e., the number of fission of the reactor core per unit time. The nuclei undergo fission after absorbing a neutron, and the energy released per fission is a constant for a defined fissile nucleus. The fissionable nuclei are nuclei which can cause fission reaction by neutron bombardment with any energy, have a large thermal neutron fission section, can initiate nuclear fission by thermal neutron bombardment with lower energy, and generate new neutrons to continue to cause nuclear fission, thereby forming a chain reaction.
The energy released by each fission of the atomic nucleus is E f, the fission rate is F, and the power P of the reactor core is:
P＝EfF (1)
considering only thermal neutron induced uranium-235 nuclear fission,
Ef＝200MeV＝3.2×10-11J
At this time, the core power of the reactor is
In the above formula, Σ f is the macroscopic fission section of the core, macroscopic fission section refers to the section of the macroscopic scale atom where the fission reaction occurs, V is the volume of the core,The average thermal neutron flux density of the core is the neutron flux density, i.e., the number of neutrons passing through a unit area perpendicular to the direction of neutron motion per unit time.
By monitoring the thermal neutron flux density with the ionization chamber, the output current of the ionization chamber can be obtained as follows:
i＝φε (3)
In the above formula, i is the output current (A) of the ionization chamber, ε is the thermal neutron sensitivity (A/n cm -2·s-1) of the ionization chamber, and φ is the thermal neutron flux density (n cm -2·s-1) monitored by the ionization chamber.
In the prior art, only the current value monitored by an ionization chamber can be measured, and the core power of a critical device cannot be directly measured. Based on the above-described problems, an embodiment of the present invention provides a method of determining core power of a critical device, which calculates a relationship coefficient by a relationship between core power of the critical device and thermal neutron flux density at a predetermined location, and by using the relationship coefficient, core power of the critical device can be obtained in real time from a current value at the predetermined location measured by an ionization chamber.
In the embodiment of the present invention, it can be understood that the simulation device is a physical model of the critical device, and each parameter of the simulation device is a theoretical value of each parameter of the critical device counted or calculated by the monte carlo method.
An embodiment of the present invention provides a method of determining core power of a critical device, fig. 1 is a schematic flow chart of a method of determining core power of a critical device according to an embodiment of the present invention, see fig. 1, the method comprising:
Step 10, simulating a critical device by using a Monte Carlo method to obtain a simulation device of the critical device, and representing the relative reactor core power of the simulation device;
Step 20, representing the relative thermal neutron flux density of a preset position by using a Monte Carlo method, wherein the preset position is arranged outside the reactor core of the critical device, and the relative thermal neutron flux density of the preset position is the relative thermal neutron flux density of neutrons generated by the reactor core;
Step 30, representing the relation between the core power of the critical device and the thermal neutron flux density of the preset position according to the relative core power of the simulation device and the relative thermal neutron flux density of the preset position;
Step 40, simulating the operation process of the simulation device by using a Monte Carlo method to calculate the relative thermal neutron flux density at a preset position and the relative core power of the simulation device, and calculating a relation coefficient according to the relation between the core power of the critical device and the thermal neutron flux density at the preset position;
Step 50, starting a critical device;
step 60, measuring the thermal neutron flux density at a preset position, and determining the core power of the critical device according to the relation coefficient.
In the embodiment of the invention, the relation between the core power of the critical device and the thermal neutron flux density of the preset position is firstly represented through the steps 10, 20 and 30, and in the steps 10, 20 and 30, only the logic relation among the parameters of the critical device is represented, and calculation is not needed. After obtaining the relationship between the core power of the critical device and the thermal neutron flux density at the predetermined position, the operation process of the critical device is simulated in step 40, each parameter value of the critical device is calculated, and the calculated result is substituted into the relationship between the core power of the characterized critical device and the thermal neutron flux density at the predetermined position to calculate the relationship coefficient. And starting the critical device to enable the critical device to normally operate, and determining the core power of the critical device in real time according to the measured thermal neutron flux density and the relation coefficient at the preset position.
Alternatively, a person skilled in the art can simulate a critical device of a certain type by using a Monte Carlo method to obtain a relation coefficient, and any critical device included in the certain type can be actually operated to determine the core power of the critical device; one skilled in the art can also simulate a specific critical device by using the monte carlo method to obtain a relationship coefficient, so that the specific critical device actually operates, and the core power of the specific critical device is determined.
The monte carlo method (Monte Carlo method), also called a statistical simulation method, is a very important numerical calculation method guided by a probability statistical theory, which is proposed in the middle of the forty-th century due to the development of science and technology and the invention of an electronic computer. Refers to a method of solving many computational problems using random numbers or more commonly pseudo-random numbers. When the solved problem is the probability of occurrence of a certain random event or the expected value of a certain random variable, estimating the probability of the random event according to the occurrence frequency of the event by a certain experimental method, or obtaining certain digital characteristics of the random variable and taking the digital characteristics as the solution of the problem. The monte carlo process can be divided into three main steps: constructing or describing a probability process, implementing sampling from known probability distributions, establishing various estimators.
According to the method for determining and understanding the core power of the device, provided by the embodiment of the invention, the relation between the core power of the critical device and the thermal neutron flux density at the preset position is obtained by simulating the critical device, and when the critical device is actually measured, only the thermal neutron flux density at the preset position is needed to be measured, so that the real-time power of the critical device is obtained, and the core power of the critical device can be continuously measured, so that the requirement that an experiment operator knows the power of the critical device in real time is met, and the critical device is ensured to safely and stably operate under different experimental conditions.
In step 10, the critical device is simulated by using the monte carlo method, which specifically includes:
The fuel element, moderator, and reflective layer in the critical device were simulated using the monte carlo method, wherein the fuel element included a fuel pellet and a cavity. When the critical device is simulated by using the Monte Carlo method, the obtained simulation device is unified with the actual situation, so that the calculation result is true and reliable. Alternatively, the core vessel and other structural components in the critical plant may also be modeled, such as the space of the core gap.
The relative core power of the simulated critical device in step 10 is an equivalent value of the core power generated by a neutron causing the critical device to fission. In step 20, optionally, the ionization chamber at the predetermined location, which may be a gamma compensated ionization chamber, a long neutron ionization chamber, or a fission ionization chamber, may be characterized to obtain a relative thermal neutron flux density at the predetermined location.
FIG. 2 is a schematic flow chart of step 30 of the method of determining core power of a critical device according to FIG. 1, see FIG. 2. In step 30, characterizing the relationship of core power of a critical device and thermal neutron flux density at a predetermined location as a function of the relative core power of the simulated device and the relative thermal neutron flux density at the predetermined location comprises:
Step 31, characterizing the core power of the critical device according to the relative core power of the simulation device;
Step 32, characterizing the relative thermal neutron flux density ratio of the simulation device, wherein the relative thermal neutron flux density ratio of the simulation device is the ratio of the relative thermal neutron flux density of the preset position to the relative thermal neutron flux density of the reactor core;
Step 33, characterizing the relationship between the core power of the critical device and the thermal neutron flux density at the predetermined location based on the core power of the critical device and the relative thermal neutron flux density ratio of the simulation device.
Further, fig. 3 is a schematic flow chart of step 31 of the method of determining core power of a critical device shown in fig. 2, referring to fig. 3, the core power of the critical device in step 31 is calculated by:
Step 311, dividing the fuel element into at least one fuel cell;
step 312, characterizing the relative fission reaction rate of the fuel cell by using a Monte Carlo method;
Step 313, characterizing relative core power, which is an equivalent value of core power generated by a neutron induced threshold device fission, based on the relative fission reaction rate;
The relative core powers of the simulation devices are characterized using steps 311, 312 and 313, and then in step 314, the core powers of the critical devices are characterized based on the relative core powers.
In particular, the relative fission reaction rate of the fuel pellets within the ith fuel element of the analog device of the critical device is characterized using the monte carlo method.
In the above formula (4), R i is the relative fission reaction rate of the ith fuel cell, V i is the fuel volume of the ith fuel cell, n is the number of fuel cells into which the fuel element is divided,For the total fission reaction rate of all fuel cells, E f is the energy released by each fission of the fuel cell, p is the relative core power, which is the equivalent of the core power generated by a neutron causing the critical device to fission.
Further, the relative thermal neutron flux density of the core in step 32 is characterized by the steps of:
the relative thermal neutron flux density of the fuel pellets, the relative thermal neutron flux density of the cavity and the relative thermal neutron flux density of the moderator are respectively characterized by utilizing a Monte Carlo method;
The relative thermal neutron flux density of the core is characterized in terms of the relative thermal neutron flux density of the fuel pellets, the relative thermal neutron flux density of the cavity, and the relative thermal neutron flux density of the moderator, wherein the moderator may be a zirconium hydride moderator.
In the above-mentioned formula (5),For simulating the relative thermal neutron flux density of the core of the device,/>The thermal neutron flux density of the fuel pellets in the fuel element characterized by the Monte Carlo method, and V F is the volume of the fuel pellets in the fuel element; /(I)The thermal neutron flux density of the inner cavity of the fuel element, which is characterized by the Monte Carlo method, and V V is the volume of the inner cavity of the fuel element; /(I)Is the thermal neutron flux density of the moderator characterized by the Monte Carlo method, and V M is the volume of the moderator.
In step 40, the operation process of the simulation device is simulated by using the monte carlo method, which specifically includes:
And simulating a neutron transport process in the operation process of the simulator by using a Monte Carlo method. The physical processes and fundamental properties within the critical device are related to the operation of neutrons generated by the core of the critical device within the critical device and the spatial energy distribution of neutron density within the core. Neutrons that originally had a certain energy and direction of motion at one location will, over time, appear at another location with another energy and direction of motion due to neutron motion and its scattering collisions with nuclei. Neutrons are transported from one location, energy and direction to another, a neutron transport process.
In the study of neutron transport problems, a deterministic method may be utilized, i.e., a mathematical model built from the physical properties of the problem may be represented by one or a set of determined mathematical physical equations, such as building a neutron transport equation, and mathematically solving for the neutron transport equation. A non-deterministic method, i.e., a monte carlo method employed by an embodiment of the present invention, may also be utilized that utilizes a series of random numbers to simulate the path of neutron motion, track the history of each neutron, and analyze the information obtained. The neutron tracking direct simulation method can be selected when the neutron transport problem is studied by utilizing the Monte Carlo method, namely, a single neutron is considered, the history of random movement of the single neutron is simulated, various parameters of the single neutron in movement, such as the relative thermal neutron flux density when the single neutron runs to a preset position, or the equivalent value of core power generated by the fission of a critical device caused by the single neutron, namely, the relative core power, are recorded, the statistical result of a large number of neutrons is assumed to be the same as the statistical result of repeated simulation of a large number of single neutrons under the same condition, therefore, random test values are obtained through tracking of the history of a large number of neutrons, and an estimation is made by utilizing the statistical method.
Specifically, the relative thermal neutron flux density ratio K of the simulation device in step 32 is shown in the following equation,
The relative thermal neutron flux density ratio is a ratio of a thermal neutron flux density at a predetermined location to an average thermal neutron flux density of a core of the critical device during actual operation of the critical device. In the above-mentioned formula (6),For the relative thermal neutron flux density of the core of the simulation device, phi M is the relative thermal neutron flux density of the predetermined location characterized by the Monte Carlo method, phi is the thermal neutron flux density of the predetermined location when the critical device is actually operating,/>Is the average thermal neutron flux density of the core of the critical device.
In characterizing parameters of a simulator of a critical device using the monte carlo method, all the above counts are normalized to the case of one source particle, i.e. one neutron.
In step 31, the core power of the critical device is characterized based on the relative core powers of the simulation devices, as shown in equation (7) below,
In the above formula, N is the neutron source intensity, which is the average thermal neutron flux of the core of the critical deviceRelative thermal neutron flux density/>, to the core of the simulation deviceP is the relative core power and P is the core power of the critical device.
Specifically, in step 40, the neutron transport process of the operation of the simulator may be simulated using the Monte Carlo method to calculate the relative thermal neutron flux density at the predetermined location and the relative core power of the simulator. Further, the relative fission reaction rate of the fuel pellets in the ith fuel element of the analog device may be calculated.
Further, in step 50, the measurement of the thermal neutron flux density at the predetermined location is specifically: the thermal neutron flux density at a predetermined location is measured using an ionization chamber and a current measurement is output. Alternatively, the ionization chamber may be a gamma compensated ionization chamber, a long neutron ionization chamber, or a fission ionization chamber.
Before starting the critical device in step 50, the method further comprises:
The relationship of the core power of the critical device and the current measurement at the predetermined location is characterized according to the relationship of the core power of the critical device and the thermal neutron flux density at the predetermined location.
In the embodiment of the invention, the gamma-compensated ionization chamber can be selected to measure the thermal neutron flux density at the preset position and output the current measurement value, so that the core power of the critical device can be determined according to the relation between the determined core power of the critical device and the current measurement value at the preset position. By using the method, the core power of the critical device can be determined by reading the real-time current measurement value output by the gamma compensation ionization chamber, and the core power of the critical device can be determined in real time.
Specifically, by combining equation (6) and equation (7), the relationship between the core power P of the critical device and the current measurement I output from the ionization chamber can be obtained:
By reading the current measurements output by the ionization chamber, the real-time power of the critical device can be determined according to equation (8) above.
Alternatively, in step 40, a relationship coefficient Q may be calculated based on the relationship between core power of the critical device and current measurements at the predetermined location, and based on the calculated relative thermal neutron flux density at the predetermined location, phi M, of the operational process of the simulation device and the calculated relative fission reaction rate, R i, of the fuel pellets within the ith fuel element of the simulation device.
Further, the core power of the critical device is determined based on the relationship coefficient Q. The method is used for determining the core power of the critical device, so that the calculation steps are reduced, and the core power of the critical device can be rapidly determined according to the current measured value at the preset position.
In step 60, after the critical device is started to operate, a neutron flux density field which changes with time is formed outside a reactor core of the critical device, the neutron flux density field comprises a thermal neutron flux density field which changes with time, a gamma compensation ionization chamber is arranged at a preset position, the gamma compensation ionization chamber at the preset position measures the thermal neutron flux density at the position, a current signal is output to represent the magnitude of the thermal neutron flux density, and optionally, a power measuring device which receives the current signal output by the gamma compensation ionization chamber and converts the current signal into a current measurement value is further included.
The method for determining the core power of the critical device provided by the embodiment of the invention can determine the core power of the critical device in real time by theoretically calculating data and combining the measured current measured value of the preset position. The real-time power of the critical device is displayed by theoretical means, other equipment or other experimental measurement is not needed, the workload is simplified, and the working efficiency is improved.
An embodiment of the present invention provides an apparatus for determining core power of a critical apparatus, and fig. 4 is a schematic structural view of an apparatus for determining core power of a critical apparatus according to an embodiment of the present invention, referring to fig. 4, the apparatus including: a simulation unit 100 for simulating the critical device to obtain a simulation device of the critical device, and representing the relative core power of the simulation device; a characterization unit 200 for characterizing the relative thermal neutron flux density at a predetermined location and determining the relationship between the core power of the critical device and the thermal neutron flux density at the predetermined location based on the relative core power of the simulation device and the relative thermal neutron flux density at the predetermined location; a measuring section 300 for measuring a thermal neutron flux density at a predetermined position; a calculating unit 400 for calculating a relationship coefficient from a relationship between the core power of the critical device and the thermal neutron flux density at a predetermined position.
The simulation section 100 is specifically used to simulate a fuel element, a moderator, and a reflective layer in a critical device, where the fuel element includes a fuel pellet and a cavity.
The characterization part 200 is also used for characterizing the core power of the critical device according to the relative core power of the simulation device.
The characterization part 200 is further used for characterizing the relative core power of the simulation device, and specifically includes:
the relative fission reaction rate of at least one fuel cell in the fuel element is characterized, and the relative core power is characterized in terms of the relative fission reaction rate, the relative core power being an equivalent value of core power generated by a neutron causing the critical device to fission.
The characterization unit 200 is further configured to characterize a relative thermal neutron flux density ratio of the simulation device, where the relative thermal neutron flux density ratio is a ratio of a relative thermal neutron flux density at a predetermined location to a relative thermal neutron flux density of the core.
The characterization part 200 is further configured to characterize the relative thermal neutron flux density of the core according to the relative thermal neutron flux density of the fuel pellets, the relative thermal neutron flux density of the cavity, and the relative thermal neutron flux density of the moderator.
The characterization part 200 is used for respectively counting the relative thermal neutron flux density of the fuel core block, the relative thermal neutron flux density of the cavity and the relative thermal neutron flux density of the moderator by the characterization part 200.
The characterization unit 200 is further configured to characterize a relationship between the core power of the critical device and the thermal neutron flux density at the predetermined location according to a ratio of the core power of the critical device to the thermal neutron flux density of the simulation device.
The characterization unit 200 is further configured to characterize a relationship between the core power of the critical device and the current measurement value at the predetermined location according to the relationship between the core power of the critical device and the thermal neutron flux density at the predetermined location.
The simulation unit 100 is also used to simulate the operation of the simulation device, and in particular, may simulate the neutron transport process during the operation of the simulation device.
The calculating part 400 is also used to calculate the relative thermal neutron flux density at a predetermined location and the relative core power of the simulation device.
The measuring unit 300 is also configured to output a current measurement value after measuring the thermal neutron flux density at a predetermined position.
Fig. 5 is a schematic structural view of a measuring part 300 according to an embodiment of the present invention, and alternatively, the measuring part 300 may be a gamma compensation ionization chamber. The ionization chamber may be used as a mid-range detector. When the nuclear reaction occurs in the critical device core, a large amount of gamma rays are generated, so that a gamma field exists in and around the critical device core. There is also an effect of gamma rays at the location where the detector is arranged. The gamma rays pass through the detector, and secondary electrons are generated by Compton scattering, photoelectric effect or electron pair effect. The generated secondary electrons ionize the gas molecules to generate ions and thus generate an output current, which can cause errors in the measurement of neutron flux by the detector. When the neutron flux is high, the effect of gamma rays is negligible. The effect of gamma rays tends to be pronounced when the neutron flux is low, where an ionization chamber is required to eliminate or reduce the effect of gamma rays on the neutron flux measurement. Fig. 3 is a schematic diagram of an ionization chamber according to one embodiment of the present invention, referring to fig. 3, the ionization chamber includes a positive voltage cavity 301 and a negative voltage cavity 302, the positive voltage cavity 301 being provided with a boron coating 303 for receiving neutrons and generating an electric current. A collecting electrode 304 is arranged between the positive voltage cavity 301 and the negative voltage cavity 302, the collecting electrode 304 receives gamma rays and generates current, the positive voltage cavity 301 is connected with positive voltage, the negative voltage cavity 302 is connected with negative voltage, and the collecting electrode 304 is grounded.
When neutrons and gamma rays are incident on the ionization chamber simultaneously, the current generated by the positive voltage cavity 301 includes both neutron generated current and gamma ray generated current, and the current generated by the negative voltage cavity 302 includes only gamma ray generated current. Since the positive voltage cavity 301 is connected to the positive voltage and the negative voltage cavity 302 is connected to the negative voltage, the current generated in the positive voltage cavity 301 is opposite to the current generated in the negative voltage cavity 302. The ionization chamber outputs a current that is the sum of the currents generated by the positive voltage cavity 301 and the negative voltage cavity 302. The current output by the ionization chamber is the current generated by receiving neutrons and is only related to neutron flux, so that the influence of gamma rays is eliminated.
Embodiments of the present invention provide a computer readable storage medium comprising instructions that when run on a computer cause the computer to perform the method of determining core power of a critical device provided by embodiments of the present invention.
The various components in the various embodiments of the invention may be integrated in a single processing unit or the various components may be physically present separately. The integrated components may be realized in hardware or in software functions.
The components, if implemented in the form of software functional components and sold or used as a stand-alone product, may be stored in a computer-readable storage medium. Based on such understanding, the present invention may implement all or part of the flow of the method of the above embodiment, or may be implemented by instructing related hardware by a computer program, where the computer program may be stored in a computer readable storage medium, and the computer program may implement the steps of each of the method embodiments described above when executed by a processor. Wherein the computer program comprises computer program code which may be in source code form, object code form, executable file or some intermediate form etc. The computer readable medium may include: any entity or device capable of carrying the computer program code, a recording medium, a U disk, a removable hard disk, a magnetic disk, an optical disk, a computer Memory, a Read-Only Memory (ROM), a random access Memory (RAM, random Access Memory), an electrical carrier signal, a telecommunications signal, a software distribution medium, and so forth.
It should also be noted that, in the embodiments of the present invention, the features of the embodiments of the present invention and the features of the embodiments of the present invention may be combined with each other to obtain new embodiments without conflict.
The present invention is not limited to the above embodiments, but the scope of the invention is defined by the claims.
Claims (4)
1.A method of determining core power of a critical device, comprising:
simulating a critical device by using a Monte Carlo method to obtain a simulation device of the critical device, and representing the relative reactor core power of the simulation device;
characterizing the relative thermal neutron flux density of a preset position by using a Monte Carlo method, wherein the preset position is arranged outside a reactor core of the critical device, and the relative thermal neutron flux density of the preset position is the relative thermal neutron flux density of neutrons generated by the reactor core;
Characterizing a relationship between core power of the critical device and thermal neutron flux density at the predetermined location based on the relative core power of the simulation device and the relative thermal neutron flux density at the predetermined location;
Simulating the operation process of the simulation device by using a Monte Carlo method to calculate the relative thermal neutron flux density of the preset position and the relative core power of the simulation device, and calculating a relation coefficient according to the relation between the core power of the critical device and the thermal neutron flux density of the preset position;
starting the critical device;
measuring the thermal neutron flux density of the preset position, and determining the core power of the critical device according to the relation coefficient;
Wherein, utilize monte carlo method to simulate critical device, include: simulating a fuel element, a moderator and a reflective layer in the critical device using a monte carlo method, wherein the fuel element comprises a fuel pellet and a cavity;
characterizing the relationship between the core power of the critical device and the thermal neutron flux density at the predetermined location based on the relative core power of the simulation device and the relative thermal neutron flux density at the predetermined location comprises:
Characterizing the core power of the critical device based on the relative core powers of the simulation devices;
Characterizing a relative thermal neutron flux density ratio of the simulation device, the relative thermal neutron flux density ratio of the simulation device being a ratio of the relative thermal neutron flux density of the predetermined location to the relative thermal neutron flux density of the core;
characterizing a relationship between core power of the critical device and thermal neutron flux density at the predetermined location based on a ratio of core power of the critical device to a relative thermal neutron flux density of the simulation device;
the relative core power of the simulation device is characterized by the following steps:
Dividing the fuel element into at least one fuel cell;
Characterizing the relative fission reaction rate of the fuel cell using a monte carlo method;
characterizing relative core power, based on said relative fission reaction rate, said relative core power being an equivalent value of core power generated by a neutron causing fission of said critical device;
the relative thermal neutron flux density of the core is characterized by the steps of:
characterizing the relative thermal neutron flux density of the fuel pellets, the relative thermal neutron flux density of the cavity, and the relative thermal neutron flux density of the moderator, respectively, using a monte carlo method;
Characterizing the relative thermal neutron flux density of the core based on the relative thermal neutron flux density of the fuel pellets, the relative thermal neutron flux density of the cavity, and the relative thermal neutron flux density of the moderator;
Measuring the thermal neutron flux density at the predetermined location includes: measuring a thermal neutron flux density at the predetermined location using an ionization chamber and outputting a current measurement;
Characterizing a relationship of core power of the critical device and current measurements at the predetermined location based on the relationship of core power of the critical device and thermal neutron flux density at the predetermined location;
the relation between the core power P of the critical device and the current measured value I of the preset position is as follows:
；
Wherein the said Is the average thermal neutron flux density of the core of the critical device, wherein phi is the thermal neutron flux density of a preset position when the critical device is actually operated, and/>For the relative thermal neutron flux density of the core of the simulator, the E f is the energy released by each fission of the fuel cells, n is the number of the fuel cells divided into the fuel cells, V i is the fuel volume of the ith fuel cell, k is the relative thermal neutron flux density ratio of the simulator, and ε is the thermal neutron sensitivity of the ionization chamber;
calculating a relationship coefficient according to the relationship between the core power of the critical device and the thermal neutron flux density of the predetermined location, including:
Calculating a relationship coefficient Q according to the relation between the core power of the critical device and the current measurement value of the preset position of the simulation device and according to the relative thermal neutron flux density phi M of the preset position of the simulation device and the relative fission reaction rate R i of the fuel core block in the ith fuel element of the simulation device:
；
Determining core power of the critical device according to the relationship coefficient, including: and determining the core power of the critical device according to the relation coefficient and the current measured value.
2. The method of claim 1, wherein simulating the operation of the simulation device using a monte carlo method comprises:
and simulating a neutron transport process in the operation process of the simulator by using a Monte Carlo method.
3. An apparatus for determining core power of a critical apparatus, comprising:
a simulation unit (100) for simulating a critical device using a Monte Carlo method, obtaining a simulation device of the critical device, and characterizing the relative core power of the simulation device; the simulation part (100) is also used for simulating the operation process of the simulation device by using a Monte Carlo method;
A characterization unit (200) for characterizing the relative thermal neutron flux density at a predetermined location by using a monte carlo method, and for characterizing the relationship between the core power of the critical device and the thermal neutron flux density at the predetermined location based on the relative core power of the simulation device and the relative thermal neutron flux density at the predetermined location; the preset position is arranged outside the reactor core of the critical device, and the relative thermal neutron flux density of the preset position is the relative thermal neutron flux density of neutrons generated by the reactor core;
a measuring section (300) for measuring a thermal neutron flux density at the predetermined position after the critical device is started;
A calculation unit (400) for calculating the relative thermal neutron flux density at the predetermined location and the relative core power of the simulation device, and for calculating a relationship coefficient from the relationship between the core power of the critical device and the thermal neutron flux density at the predetermined location; determining the core power of the critical device according to the relation coefficient;
wherein the simulation part (100) is specifically used for simulating a fuel element, a moderator and a reflecting layer in the critical device, wherein the fuel element comprises a fuel pellet and a cavity;
The characterization part (200) is further used for characterizing the relative core power of the simulation device, and specifically comprises: dividing the fuel element into at least one fuel cell; characterizing a relative fission reaction rate of at least one fuel cell in the fuel element using a monte carlo method, and characterizing a relative core power as a function of the relative fission reaction rate, the relative core power being an equivalent value of core power generated by a neutron causing fission of the critical device;
The characterization part (200) is further configured to: characterizing core power of the critical device from the relative core power of the simulation device; characterizing a relative thermal neutron flux density ratio of the simulation device, the relative thermal neutron flux density ratio being a ratio of the relative thermal neutron flux density of the predetermined location and the relative thermal neutron flux density of the core; and characterizing a relationship of core power of the critical device and thermal neutron flux density at the predetermined location based on a ratio of core power of the critical device to a relative thermal neutron flux density of the simulation device;
The characterization part (200) is further configured to: characterizing the relative thermal neutron flux density of the fuel pellets, the relative thermal neutron flux density of the cavity, and the relative thermal neutron flux density of the moderator, respectively, using a monte carlo method; and characterizing the relative thermal neutron flux density of the core based on the relative thermal neutron flux density of the fuel pellets, the relative thermal neutron flux density of the cavity, and the relative thermal neutron flux density of the moderator;
the measuring part (300) is also used for outputting a current measured value after measuring the thermal neutron flux density at the preset position;
the characterization part (200) is further used for characterizing the relation between the core power of the critical device and the current measurement value of the preset position according to the relation between the core power of the critical device and the thermal neutron flux density of the preset position;
the relation between the core power P of the critical device and the current measured value I of the preset position is as follows:
；
Wherein the said Is the average thermal neutron flux density of the core of the critical device, wherein phi is the thermal neutron flux density of a preset position when the critical device is actually operated, and/>For the relative thermal neutron flux density of the core of the simulator, the E f is the energy released by each fission of the fuel cells, n is the number of the fuel cells divided into the fuel cells, V i is the fuel volume of the ith fuel cell, k is the relative thermal neutron flux density ratio of the simulator, and ε is the thermal neutron sensitivity of the ionization chamber;
calculating a relationship coefficient according to the relationship between the core power of the critical device and the thermal neutron flux density of the predetermined location, including:
Calculating a relationship coefficient Q according to the relation between the core power of the critical device and the current measurement value of the preset position of the simulation device and according to the relative thermal neutron flux density phi M of the preset position of the simulation device and the relative fission reaction rate R i of the fuel core block in the ith fuel element of the simulation device:
；
Determining core power of the critical device according to the relationship coefficient, including: and determining the core power of the critical device according to the relation coefficient and the current measured value.
4. A computer readable storage medium comprising instructions which, when run on a computer, cause the computer to perform the method of any of claims 1-2.
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