CN114169164A - Method and device for determining the core power of a critical device - Google Patents
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Abstract
The embodiment of the invention discloses a method and a device for determining the 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 reactor core power of the simulation device; characterizing the relative thermal neutron flux density of a predetermined position by using a Monte Carlo method; according to the relative core power of the simulation device and the relative thermal neutron flux density of a preset position, representing the relation between the core power of the critical device and the thermal neutron flux density of the preset position; 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 reactor core power of the simulation device, and calculating a relation coefficient according to the relation between the reactor 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 which is 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 nuclear fuel and other materials and providing basis for checking theoretical calculation. The critical apparatus can maintain controlled chain reactions at low power levels and provide conditions for studying core placement and composition.
When the critical device runs, experimenters need to know the power of the critical device in real time, so that the critical device can run safely and stably under different experimental conditions.
Disclosure of Invention
In view of the above, the present invention has been made to provide a method and apparatus for determining core power of a critical apparatus that overcomes or at least partially solves the above-mentioned problems.
A first aspect of embodiments of the present invention provides a method of determining core power of a critical plant, 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; the method comprises the steps of utilizing a Monte Carlo method to represent the relative thermal neutron flux density of a preset position, 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 the core power of the critical device and the thermal neutron flux density at the predetermined location according to 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 reactor core power of the simulation device, and calculating a relation coefficient according to the relation between the reactor 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 a critical device to obtain the 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; the characterization part is used for characterizing the relative thermal neutron flux density of a preset position and further 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 calculation unit for calculating a relation coefficient based on a relation between the core power of the critical apparatus and the thermal neutron flux density at the predetermined position.
A third aspect of the present invention provides a computer-readable storage medium comprising instructions which, when executed on a computer, cause the computer to perform the method of determining core power of a critical apparatus as provided in the first aspect of the embodiments of the present invention.
Drawings
Other objects and advantages of the present invention will become apparent from the following description of the invention which refers to the accompanying drawings, and may assist in a comprehensive understanding of the invention.
FIG. 1 is a schematic flow diagram of a method of determining core power of a critical apparatus according to one embodiment of the present invention;
FIG. 2 is a schematic flow chart diagram of step 30 of the method of determining core power of a critical plant shown in FIG. 1;
FIG. 3 is a schematic flow chart diagram according to step 31 of the method of determining core power of a critical apparatus shown in FIG. 2;
FIG. 4 is a schematic diagram of the structure of an apparatus for determining core power of a critical apparatus according to an embodiment of the present invention;
fig. 5 is a schematic structural view of a measuring part 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 structure or function 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 illustrating preferred embodiments and are not intended to limit the invention itself. The drawings do not show every aspect of the described embodiments and do not limit the scope of the invention.
In the figure, 100 is a simulation part, 200 is a characterization part, 300 is a measurement part, 301 is a positive voltage chamber, 302 is a negative voltage chamber, 303 is boron-coated, 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 described clearly and completely with reference to the accompanying drawings of the embodiments of the present invention. It should be apparent that the described embodiment is one embodiment of the invention, and not all embodiments. All other embodiments, which can be derived by a person skilled in the art from the described embodiments of the invention without any inventive step, are within the scope of protection of the invention.
Unless defined otherwise, technical or scientific terms used herein shall have 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 which is 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 nuclear fuel and other materials and providing basis for checking theoretical calculation. The fissionable materials are contained in the critical device, so that the controllable chain reaction can be maintained at a low power level, and conditions are provided for researching the arrangement and the composition of the reactor core.
The core power of a critical plant can be obtained from the core's rate of fission, i.e., the number of reactor core fissures per unit time. The nucleus will undergo fission after absorbing a neutron, and the energy released by each fission is a constant for a certain fissile nucleus. The easy fission nucleus is the nucleus which can cause the fission reaction by the bombardment of neutrons with any energy, has a large thermal neutron fission cross section, can cause the nuclear fission by the bombardment of the thermal neutrons with lower energy, and generates new neutrons to continue causing the nuclear fission to form a chain reaction.
The energy released by each fission of the nucleus is EfAnd the fission rate is F, the power P of the reactor core is:
P＝EfF (1)
considering only the uranium-235 nuclear fission caused by thermal neutrons,
Ef＝200MeV＝3.2×10-11J
the core power of the reactor is at the time
In the above formula, sigmafIs a macroscopic fission cross section of the reactor core, the macroscopic fission cross section refers to a cross section of macroscopic scale atoms for fission reaction, V is the volume of the reactor core,
the average thermal neutron flux density of the reactor core is the neutron flux density, namely the number of neutrons passing through a unit area perpendicular to the neutron motion direction in 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, and ε is the thermal neutron sensitivity (A/n cm) of the ionization chamber-2·s-1) Phi is the thermal neutron flux density (n cm) monitored by the ionization chamber-2·s-1)。
In the prior art, only the current value monitored by the ionization chamber can be measured, and the reactor core power of the critical device cannot be directly measured. In view of the above problems, embodiments of the present invention provide a method of determining core power of a critical apparatus, which calculates a relation coefficient by a relation between the core power of the critical apparatus and a thermal neutron flux density at a predetermined position, and can obtain the core power of the critical apparatus in real time from a current value at the predetermined position measured by an ionization chamber using the relation coefficient.
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 apparatus, and fig. 1 is a schematic flow chart of the method of determining core power of a critical apparatus according to an embodiment of the present invention, referring to fig. 1, the method including:
and step 60, measuring the thermal neutron flux density at a preset position, and determining the reactor 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 at the preset position is firstly characterized through the steps 10, 20 and 30, and in the steps 10, 20 and 30, only the logical relation among the parameters of the critical device needs to be characterized, and no calculation is needed. After the relation between the core power of the critical device and the thermal neutron flux density at the preset position is obtained, the operation process of the critical device is simulated through step 40, each parameter value of the critical device is calculated, the calculated result is substituted into the relation between the core power of the critical device and the thermal neutron flux density at the preset position, and the relation coefficient is calculated. 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 of the preset position.
Alternatively, one skilled in the art can simulate a certain type of critical device by the monte carlo method to obtain a relationship coefficient, so that any critical device included in the type can be actually operated to determine the core power; one skilled in the art can also simulate a specific critical device by the monte carlo method to obtain a relationship coefficient, so that the specific critical device actually operates to determine the core power.
The Monte Carlo method (also called statistical simulation method) is a very important numerical calculation method which is proposed in the middle of the fortieth century by the development of scientific technology and the invention of electronic computers and is guided by the probability statistical theory. Refers to a method of solving many computational problems using random numbers or more commonly pseudo-random numbers. When the problem to be solved is the probability of occurrence of a random event or the expected value of a random variable, the probability of the random event is estimated by using the frequency of occurrence of the event through an experimental method, or some digital characteristics of the random variable are obtained and used as the solution of the problem. The monte carlo method can be divided into three main steps: constructing or describing a probabilistic process, enabling 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, when the critical device is actually measured, the thermal neutron flux density at the preset position is only required to be measured, so that the real-time power of the critical device is obtained, the core power of the critical device can be continuously measured, the requirement of experiment operators on knowing the power of the critical device in real time is met, and the safe and stable operation of the critical device under different experiment conditions is guaranteed.
In step 10, simulating a critical device by using a monte carlo method, specifically comprising:
the fuel elements, moderator and reflective layer in the critical apparatus were simulated using the monte carlo method, wherein the fuel elements included fuel pellets and cavities. When the Monte Carlo method is used for simulating the critical device, the obtained simulation device is kept unified with the actual situation, so that the calculation result is real and credible. Alternatively, the core vessel and other structural components in the critical plant may also be modeled, such as the space for the core clearance.
The relative core power of the simulated critical apparatus in step 10 is an equivalent of the core power generated by the fission of the critical apparatus caused by neutrons. In step 20, optionally, the ionization chamber at the predetermined location, which may be a gamma compensation ionization chamber, a long neutron ionization chamber, or a fission ionization chamber, may be characterized to obtain the relative thermal neutron flux density at the predetermined location.
Fig. 2 is a schematic flow chart of step 30 of the method for determining the core power of a critical apparatus according to fig. 1, and referring to fig. 2, in step 30, the relationship characterizing the core power of the critical apparatus and the thermal neutron flux density at a predetermined position according to the relative core power of the simulation apparatus and the relative thermal neutron flux density at the predetermined position comprises:
and step 33, representing the relation between the core power of the critical device and the thermal neutron flux density at a preset position according to the ratio between the core power of the critical device and the relative thermal neutron flux density of the simulation device.
Further, fig. 3 is a schematic flow chart of step 31 of the method for determining the core power of the critical apparatus shown in fig. 2, and referring to fig. 3, the core power of the critical apparatus in step 31 is calculated by:
the relative core power of the simulation plant is characterized by steps 311, 312 and 313, and the core power of the critical plant is further characterized in step 314 based on the relative core power.
In particular, the relative fission reactivity of fuel pellets within the ith fuel element of the simulation setup of the critical setup was characterized using the monte carlo method.
In the above formula (4), RiIs the relative fission reactivity, V, of the ith fuel celliIs 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, EfThe energy released by each fission of the fuel grid element is p, the relative core power is the equivalent value of the core power generated by the fission of the critical device caused by neutrons.
Further, the relative thermal neutron flux density of the core in step 32 is characterized by the following steps:
respectively representing the relative thermal neutron flux density of the fuel pellet, the relative thermal neutron flux density of the cavity and the relative thermal neutron flux density of the moderator by using 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, which may be a zirconium hydride moderator.
In the above-mentioned formula (5),
to simulate the relative thermal neutron flux density of the core of the plant,
thermal neutron flux density, V, of fuel pellets within a fuel element characterized by the Monte Carlo methodFIs the volume of fuel pellets within the fuel element;
thermal neutron flux density, V, of a cavity within a fuel element characterized by the Monte Carlo methodVIs the volume of the cavity within the fuel element;
thermal neutron flux density, V, of moderator characterized by the Monte Carlo methodMIs the volume of moderator.
In step 40, simulating the operation process of the simulation device by using a monte carlo method, specifically comprising:
and simulating the neutron transport process in the operation process of the simulation device by using a Monte Carlo method. The physical processes and fundamental characteristics of the critical apparatus are related to the operation of neutrons generated by the core of the critical apparatus in the critical apparatus and the spatial energy distribution of neutron density in the core. Due to neutron motion and its scattering collisions with atomic nuclei, neutrons that originally have a certain energy and direction of motion at a certain location will, over some time, emerge at another location with another energy and direction of motion. Neutrons are transported from one location, energy and direction to another location, energy and direction, and this process is a neutron transport process.
In studying neutron transport problems, deterministic methods may be used, i.e., a mathematical model established from the physics of the problem may be represented by a defined mathematical equation or set of mathematical physics equations, e.g., a neutron transport equation may be established and an accurate or approximate solution may be mathematically solved for the neutron transport equation. A non-deterministic method, i.e., the monte carlo method employed by embodiments of the present invention, can also be utilized, which utilizes a series of random numbers to simulate the path of neutron motion, tracks the history of each neutron, and analyzes the information obtained. The method for researching the neutron transport problem by using the Monte Carlo method can adopt a neutron tracking direct simulation method, namely, a single neutron is considered, the history of random motion of the single neutron is simulated, various parameters in motion of the single neutron, such as the relative thermal neutron flux density when the single neutron is operated to a preset position, or the equivalent value of the reactor core power generated by the fission of a critical device caused by the single neutron, namely the relative reactor 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 the large number of single neutrons under the same condition, therefore, the random test value is obtained by tracking the history of the large number of neutrons, and the estimation quantity is made by using the statistical method.
Specifically, the relative thermal neutron flux density ratio K of the simulation apparatus in step 32 is shown in the following formula,
during the actual operation of the critical device, the relative thermal neutron flux density ratio is the thermal neutron flux density at a predetermined position to the core of the critical deviceRatio of average thermal neutron flux density. In the above-mentioned formula (6),
to simulate the relative thermal neutron flux density, phi, of the core of the plantMThe relative thermal neutron flux density of a preset position is characterized by using a Monte Carlo method, phi is the thermal neutron flux density of the preset position when the critical device is actually operated,
is the average thermal neutron flux density of the core of the critical plant.
When characterizing the parameters of the simulation device 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 plant is characterized according to the relative core power of the simulation plant, as shown in the following formula (7),
in the above formula, N is the neutron source intensity, which is the average thermal neutron flux of the core of the critical apparatus
Relative thermal neutron flux density to core of simulation device
P is the relative core power, P is the core power of the critical plant.
Specifically, in step 40, a neutron transport process of the operation process of the simulation apparatus may be simulated using a monte carlo method to calculate the relative thermal neutron flux density at the predetermined location and the relative core power of the simulation apparatus. Further, the relative fission reactivity of the fuel pellets within the ith fuel element of the simulation device may be calculated.
Further, in step 50, measuring the thermal neutron flux density at the predetermined position specifically includes: 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 compensation ionization chamber, a long neutron ionization chamber, or a fission ionization chamber.
Before the critical device is activated in step 50, the method further comprises:
and according to the relation between the core power of the critical device and the thermal neutron flux density at the preset position, representing the relation between the core power of the critical device and the current measurement value at the preset position.
In the embodiment of the invention, the gamma compensation 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, combining equation (6) and equation (7), the relationship between the core power P of the critical device and the current measurement value I output by 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.
Optionally, in step 40, the relative thermal neutron flux density φ at the predetermined position can be calculated according to the relation between the core power of the critical device and the current measurement value at the predetermined position and according to the operation process of the simulation deviceMAnd simulating a relative fission reactivity R of fuel pellets in an ith fuel element of the deviceiAnd calculating a relation coefficient Q.
Further, the core power of the critical device is determined according to the relation 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 quickly determined according to the current measurement value of the preset position.
In step 60, after the critical apparatus is started to operate, a neutron flux density field which changes with time is formed outside a core of the critical apparatus, the neutron flux density field comprises a thermal neutron flux density field which changes with time, the 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, and a current signal is output to represent the magnitude of the thermal neutron flux density.
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 through theoretical calculation data and combined with the measured current measurement value of the preset position. The real-time power of the critical device is displayed by a 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 the apparatus for determining core power of a critical apparatus according to an embodiment of the present invention, referring to fig. 4, the apparatus including: the simulation part 100 is used for simulating the critical device to obtain a simulation device of the critical device and representing the relative reactor core power of the simulation device; the characterization part 200 is used for characterizing the relative thermal neutron flux density of the preset position and determining 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 300 for measuring a thermal neutron flux density at a predetermined position; the calculation unit 400 is configured to calculate a relation coefficient from a relation between the core power of the critical apparatus and the thermal neutron flux density at a predetermined position.
The simulation section 100 is specifically configured to simulate fuel elements, including fuel pellets and cavities, moderator and reflective layers in a critical apparatus.
The characterizer 200 is also used to characterize the core power of the critical plant based on the relative core powers of the simulated plants.
The characterization part 200 is further used for characterizing the relative core power of the simulation apparatus, and specifically includes:
the method includes the steps of characterizing a relative fission reactivity of at least one fuel cell in the fuel element, and characterizing a relative core power from the relative fission reactivity, the relative core power being an equivalent of the core power generated by neutrons causing fission in the critical device.
The characterization part 200 is also used for characterizing a relative thermal neutron flux density ratio of the simulation device, wherein the relative thermal neutron flux density ratio is a ratio of a relative thermal neutron flux density of a preset position and a relative thermal neutron flux density of the core.
The characterization part 200 is also used to characterize the relative thermal neutron flux density of the core from 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.
And a characterization part 200, wherein the characterization part 200 is used for respectively counting the relative thermal neutron flux density of the fuel pellet, the relative thermal neutron flux density of the cavity and the relative thermal neutron flux density of the moderator.
The characterization part 200 is further used for characterizing the relation between the core power of the critical device and the thermal neutron flux density at a preset position according to the core power of the critical device and the relative thermal neutron flux density ratio of the simulation device.
The characterization part 200 is further configured to characterize the relationship between the core power of the critical apparatus and the current measurement value at the predetermined position according to the relationship between the core power of the critical apparatus and the thermal neutron flux density at the predetermined position.
The simulation part 100 is also used for simulating the operation process of the simulation apparatus, and particularly, can simulate the neutron transport process during the operation process of the simulation apparatus.
The calculation part 400 is also used to calculate the relative thermal neutron flux density at a predetermined position and the relative core power of the simulation apparatus.
The measuring unit 300 is also used 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 measurement part 300 according to an embodiment of the present invention, and alternatively, the measurement part 300 may be a gamma compensation ionization chamber. The ionization chamber can be used as an intermediate range detector. Since a large amount of gamma rays are generated when the critical apparatus core undergoes a nuclear reaction, gamma fields exist in and around the critical apparatus core. The detector is also located at a position where there is an influence of the gamma ray. When gamma rays pass through the detector, secondary electrons are generated by compton scattering, the photoelectric effect or the electron pair effect. The generated secondary electrons can ionize gas molecules to generate ions, and then an output current is generated, and the output current can cause errors of neutron flux measured by the detector. When the neutron flux is high, the effect of gamma rays is negligible. The effect of gamma rays tends to be significant when the neutron flux is low, and it is desirable to use an ionization chamber to eliminate or reduce the effect of gamma rays on the neutron flux measurement. Fig. 3 is a schematic structural view of an ionization chamber according to an embodiment of the present invention, and referring to fig. 3, the ionization chamber includes a positive voltage chamber 301 and a negative voltage chamber 302, the positive voltage chamber 301 being provided with 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 simultaneously emitted into the ionization chamber, the current generated by the positive voltage cavity 301 includes both the current generated by the neutrons and the current generated by the gamma rays, and the current generated by the negative voltage cavity 302 includes only the current generated by the gamma rays. Since the positive voltage chamber 301 is connected to a positive voltage and the negative voltage chamber 302 is connected to a negative voltage, the current generated by the positive voltage chamber 301 is opposite to the current generated by the negative voltage chamber 302. The current output by the ionization chamber 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 executed on a computer, cause the computer to perform a method of determining core power of a critical apparatus as provided by embodiments of the present invention.
The components in the embodiments of the present invention may be integrated into one processing unit, or may exist separately and physically. The integrated components may be implemented in the form of hardware, or may be implemented in the form of software functions.
The components, if implemented in software features and sold or used as a stand-alone product, may be stored in a computer readable storage medium. Based on such understanding, all or part of the flow in the method according to the above embodiments may be implemented by a computer program, which may be stored in a computer readable storage medium and used by a processor to implement the steps of the above embodiments of the method. Wherein the computer program comprises computer program code, which may be in the form of source code, object code, an executable file or some intermediate form, etc. The computer-readable medium may include: any entity or device capable of carrying the computer program code, recording medium, usb disk, removable hard disk, magnetic disk, optical disk, computer Memory, Read-Only Memory (ROM), Random Access Memory (RAM), electrical carrier wave signals, telecommunications signals, software distribution medium, and the like.
It should also be noted that, in the case of the embodiments of the present invention, features of the embodiments and examples may be combined with each other to obtain a new embodiment without conflict.
The above description is only an embodiment of the present invention, but the scope of the present invention is not limited thereto, and the scope of the present invention is subject to the scope of the claims.
Claims (21)
1. A method of determining core power of a critical plant, 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;
the method comprises the steps of utilizing a Monte Carlo method to represent the relative thermal neutron flux density of a preset position, 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 the core power of the critical device and the thermal neutron flux density at the predetermined location according to 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 reactor core power of the simulation device, and calculating a relation coefficient according to the relation between the reactor 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.
2. The method of claim 1, wherein simulating the critical device using a monte carlo method comprises:
a fuel element, a moderator, and a reflective layer in the critical apparatus are simulated using a monte carlo method, wherein the fuel element comprises a fuel pellet and a cavity.
3. The method of claim 2, wherein characterizing the relationship between the core power of the critical plant and the thermal neutron flux density at the predetermined location as a function of the relative core power of the simulated plant and the relative thermal neutron flux density at the predetermined location comprises:
characterizing the core power of the critical plant from the relative core power of the simulation plant;
characterizing a relative thermal neutron flux density ratio of the simulation device, wherein the relative thermal neutron flux density ratio of the simulation device is a ratio of the relative thermal neutron flux density of the predetermined position to the relative thermal neutron flux density of the reactor core;
and characterizing the relation between the core power of the critical device and the thermal neutron flux density of the preset position according to the core power of the critical device and the relative thermal neutron flux density ratio of the simulation device.
4. The method of claim 3, wherein the relative core power of the simulation plant is characterized by:
dividing the fuel element into at least one fuel cell;
characterizing relative fission reactivity of the fuel cells using a Monte Carlo method;
and characterizing relative core power according to the relative fission reactivity, wherein the relative core power is an equivalent value of the core power generated by neutrons causing the fission of the critical device.
5. The method of claim 3, wherein the relative thermal neutron flux density of the core is characterized by:
respectively 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 by using a Monte Carlo method;
characterizing a relative thermal neutron flux density of the core from 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.
6. The method of claim 1, wherein simulating the operation of the simulation device using a monte carlo method comprises:
and simulating the neutron transport process in the operation process of the simulation device by using a Monte Carlo method.
7. The method of claim 1, wherein measuring the thermal neutron flux density at the predetermined location comprises measuring the thermal neutron flux density at the predetermined location with an ionization chamber and outputting a current measurement.
8. The method of claim 7, wherein prior to calculating the relationship coefficients, further comprising:
and 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.
9. An apparatus for determining core power for a critical plant, comprising:
the simulation part (100) is used for simulating a critical device, obtaining a simulation device of the critical device and representing the relative reactor core power of the simulation device;
a characterization part (200) for characterizing the relative thermal neutron flux density of a predetermined position and further for characterizing the relation between the core power of the critical device and the thermal neutron flux density of the predetermined position according to the relative core power of the simulation device and the relative thermal neutron flux density of the predetermined position;
a measuring unit (300) for measuring the thermal neutron flux density at the predetermined position;
and a calculation unit (400) for calculating a relation coefficient from the relation between the core power of the critical apparatus and the thermal neutron flux density at the predetermined position.
10. The device according to claim 9, characterized in that the simulation section (100) is particularly adapted to simulate fuel elements, moderators and reflective layers in the critical device, wherein the fuel elements comprise fuel pellets and cavities.
11. The apparatus of claim 10, wherein the characterization unit (200) is further configured to characterize the core power of the critical plant based on the relative core power of the simulated plant.
12. The apparatus according to claim 11, characterized in that the characterization part (200) is also used for characterizing the relative core power of the simulation apparatus, in particular comprising:
characterizing a relative fission reactivity of at least one of the fuel elements, and characterizing a relative core power from the relative fission reactivity, the relative core power being an equivalent of the core power produced by a neutron causing fission of the critical device.
13. The apparatus of claim 11, wherein the characterization unit (200) is further configured to characterize a relative thermal neutron flux density ratio of the simulation apparatus, the relative thermal neutron flux density ratio being a ratio of a relative thermal neutron flux density of the predetermined location and a relative thermal neutron flux density of the core.
14. The apparatus of claim 13, wherein the characterization unit (200) is further configured to characterize 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.
15. The apparatus of claim 14, wherein the characterization unit (200) is further configured to characterize a relative thermal neutron flux density of the fuel pellets, a relative thermal neutron flux density of the cavity, and a relative thermal neutron flux density of the moderator, respectively.
16. The apparatus of claim 13, wherein the characterizer (200) is further configured to characterize the relationship between the core power of the critical apparatus and the thermal neutron flux density at the predetermined location based on the core power of the critical apparatus and the relative thermal neutron flux density ratio of the simulated apparatus.
17. The apparatus of claim 9, wherein the characterizer (200) is further configured to characterize the relationship of the core power of the critical apparatus and the current measurement at the predetermined location based on the relationship of the core power of the critical apparatus and the thermal neutron flux density at the predetermined location.
18. The apparatus according to claim 9, wherein the measuring section (300) is further configured to output a current measurement value after measuring the thermal neutron flux density at the predetermined position.
19. The device according to claim 9, characterized in that the simulation unit (100) is also adapted to simulate the operation of the simulation device.
20. The apparatus of claim 9, wherein said calculating section (400) is further configured to calculate a relative thermal neutron flux density at said predetermined location and a relative core power of said simulation apparatus.
21. A computer-readable storage medium comprising instructions that, when executed on a computer, cause the computer to perform the method of any of claims 1-8.
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