CN113609744A - Reactor core three-dimensional power rapid construction method based on Monte Care critical calculation one-step method - Google Patents
Reactor core three-dimensional power rapid construction method based on Monte Care critical calculation one-step method Download PDFInfo
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Abstract
A reactor core three-dimensional power rapid construction method based on a Monte Care critical calculation single-step method increases the number of particles generation by generation, does not calculate inactive generation, does not need to consider fission source convergence diagnosis problems, obviously improves the Monte Care critical calculation efficiency in a mode of adjusting the weight of each calculation generation in real time, and ensures that all calculation generations contribute to calculation results. Based on the Monte card critical single-step method without the inactive generation, the full-pile three-dimensional power distribution can be quickly constructed, and the highest calculation efficiency is realized.
Description
Technical Field
The invention relates to a technology in the field of reactor physics, in particular to a method for avoiding inactive generation by adjusting the particle number scale of Montgomery critical calculation, thereby improving the efficiency of Montgomery critical calculation and realizing the rapid construction of three-dimensional power of a reactor core.
Background
The three-dimensional power distribution of the reactor core is the most critical physical quantity in the research and development design of the novel reactor and can be obtained through Monte-Ka critical calculation, so the critical calculation is the most basic requirement of the physical design of the reactor. The traditional Monte Care critical algorithm divides the calculation process into an inactive generation and an active generation, namely: and obtaining stable and convergent fission source distribution through calculation of the inactive generation, and counting system information through calculation of the active generation. In which the calculation of the inactive generation is not counted, which is regarded as a waste of calculation resources. In some large-scale systems and loosely coupled systems, fission source convergence is slow, and it takes hundreds of inactive generations of calculations to achieve fission source convergence. Meanwhile, in the calculation process of the non-active generation, whether the fission source is converged is difficult to diagnose in real time, and calculation errors caused by starting statistics after the fission source is not converged are avoided by increasing the non-active generation, so that the waste of calculation resources is further caused. If the computational efficiency is too low, the three-dimensional power distribution of the reactor core cannot be quickly constructed, so that the high-fidelity physical calculation of the reactor core faces difficulty, and the research and development design of a novel reactor faces a bottleneck.
Disclosure of Invention
Aiming at the problem of low calculation efficiency of the existing Monte Care critical algorithm caused by the inactive generation process, the invention provides a reactor core three-dimensional power rapid construction method based on a Monte Care critical calculation one-step method, which can remarkably improve the Monte Care critical calculation efficiency by increasing the number of particles generation by generation and adjusting the weight of each calculation generation in real time, thereby rapidly constructing the reactor core three-dimensional power distribution and effectively supporting the reactor core physical design of a novel reactor.
The invention is realized by the following technical scheme:
the invention relates to a reactor core three-dimensional power rapid construction method based on a Monte Care critical computation one-step method, which comprises the following steps:
step 1: determining an optimal population increment sequence according to a fission source error transfer model: obtaining a generic population increment sequence from a fission source error transfer model
Wherein: i represents a calculationThe serial number of the generation is 1, 2 and 3; m is(1)Is the initial particle number, m(i)Is the number of particles of the ith generation; c is a constant associated with the model. By determining the constant c and the initial particle number m(1)The optimal population increment sequence can be determined.
The present invention has demonstrated c and m by mathematical derivation(1)The smaller the value of (A), the higher the calculation efficiency of the Monte Carlo critical calculation single-step method. In the invention, the constant c is preferably 1.0; number of initial particles m(1)The initial number of particles m required is related to the system size and the degree of loose coupling, the larger the system or the higher the degree of loose coupling(1)The larger the value is, the infinite reduction of the initial particle number m cannot be achieved for improving the calculation efficiency(1)Initial number of particles m in the present invention(1)Preferably 50, 100 or 1000.
Step 2: and determining the optimized weight of each calculation generation to the overall calculation result according to the particle number increasing sequence. Because the previous p generation fission source has larger error, the optimal weighting coefficient is obtained by minimizing the proportion of the previous p generation neutron statistical information to the total neutron statistical information on the premise of fixing the total particle number, thereby realizing the highest calculation efficiency.
The previous p generation neutron statistical information accounts for the share of the total neutron statistical information
Wherein:
information quantity of previous p generations; scumIs the total information amount; n represents the total algebra of the simulation; w is a(i)Is a weighting coefficient of the ith generation; m is(i)Is the number of particles of the ith generation; c is a constant associated with the model.
The optimized weighting coefficient
And step 3: performing Monte Care critical calculation and counting the three-dimensional power distribution of the reactor core: ensuring that the number of particles in each generation is equal to the optimal number-of-particles increasing sequence, no longer distinguishing the inactive generation from the active generation, counting the power distribution from the 1 st calculation generation, adopting the optimized weighting coefficient for each generation of weighting coefficients, and further constructing the three-dimensional power distribution of the reactor core
Wherein: i is the serial number of the calculation generation; w is a(i)Is a weighting coefficient of the ith generation; m is(i)Is the number of particles of the ith generation; r (m)(i)) Is the power response function of the ith generation, Q(i)Is the power distribution of the ith generation, and N is the total number of computations.
Technical effects
The method integrally solves the problem of low calculation efficiency of the traditional Monte Carlo critical algorithm. According to the Monte Carlo critical computation method, the inactive generation is not calculated any more, the convergence diagnosis problem of the fission source does not need to be considered, all the calculation generations are guaranteed to contribute to the calculation result, and the Monte Carlo critical computation effect is effectively improved. Based on the Monte card critical single-step method without the inactive generation, the full-pile three-dimensional power distribution can be quickly constructed, and the highest calculation efficiency is realized.
Drawings
FIG. 1 is a flow chart of the present invention;
FIG. 2 is a geometry of an embodiment VERA reactor core;
FIG. 3 is a three-dimensional power distribution diagram of an embodiment.
Detailed Description
As shown in fig. 2, for an application scenario of the non-inactive generation monte critical computation single-step method according to this embodiment, that is, a VERA reactor core, a core is uniformly divided into 15 × 15 × 10 regions 2250, and power distributions of the 2250 regions are computed by using this method, where the power distribution of the core region is shown in fig. 3.
The embodiment firstly uses an RMC program to carry out physical modeling on the model, and defines the geometric and material parameters of the model through an input card; selecting for the calculation modelThe sequence of increasing numbers of particles, namely input: m is(1)C and h, and according to m(1)C and h determine the particle number sequence m(i)(ii) a Determining a total algebra N to be calculated according to the total particle number; and performing mathematical optimization analysis on the whole model to determine the optimal weighting coefficient.
From generation 1, neutron transport simulations were performed, with the number of neutrons for each generation satisfying the increasing sequence of particle numbers.
The statistics of the power distribution is carried out from the 1 st generation, and the statistical process meets the requirements of the weighting technology.
And after N-generation calculation is completed, analyzing data and processing results to obtain the three-dimensional power distribution of the whole stack.
To highlight the advantages of the present solution over the prior art, a comparative verification is performed here. The traditional technical scheme is adopted to carry out 8 groups of calculation, and the calculation parameters are as follows: (1) m is 10, 000, n1＝100，n2＝500；(2)m＝10，000，n1＝200，n2＝500；(3)m＝100，000，n1＝100，n2＝500；(4)m＝100，000，n1＝200，n2＝500；(5)m＝500，000，n1＝100，n2＝500；(6)m＝500，000，n1＝200，n2＝500；(7)m＝1，000，000，n1＝100，n2＝500；(8)m＝1，000，000，n1＝200，n2500. Where m denotes the number of particles simulated per generation, n1The number of inactive generations of a simulation is indicated, and the number of total calculated generations of a simulation is indicated.
Meanwhile, 8 groups of calculation are carried out by using the calculation scheme, and the calculation parameters are as follows: (1) m is1＝50，c＝1.0，n＝1000；(2)m1＝50，c＝1.0，n＝5000；(3)m1＝50，c＝1.0，n＝10，000；(4)m1＝100，c＝1.0，n＝100，000；(5)m1＝100，c＝1.0，n＝1000；(6)m1＝100，c＝1.0，n＝5000；(7)m1＝100，c＝1.0，n＝10，000；(8)m1100, c 1.0, and n 100, 000. M here1Denotes an initial particle number, c denotes a constant of a sequence of increasing particle numbers, nThe number of calculated generations for the total simulation is indicated.
Using time dependent average quality factor to measure computational efficiency
Wherein: m is the number of statistical regions, T is the calculation time, RekIs the relative standard deviation of the kth region.
While the calculation efficiency is measured by using the average quality factor related to the number of particles
Wherein: m is the number of statistical regions, h is the total number of particles simulated, RekIs the relative standard deviation of the kth region.
Two initial fission sources were set for the critical calculations: (1) an initial point source at a central location of the VERA core; (2) a homogeneous source throughout the core of the VERA core. The results of the calculations are compared to table 1.
TABLE 1 comparison of the mean quality factor calculated for each group
As can be seen from table 1 and fig. 3, the method can rapidly construct the power distribution of the three-dimensional full core.
Compared with the prior art, the method has higher overall calculation efficiency. The efficiency of Monte Care critical calculation can be effectively improved, so that the three-dimensional power distribution of the reactor core is quickly constructed, and the physical design of the reactor core of the novel reactor is effectively supported.
The foregoing embodiments may be modified in many different ways by those skilled in the art without departing from the spirit and scope of the invention, which is defined by the appended claims and all changes that come within the meaning and range of equivalency of the claims are therefore intended to be embraced therein.
Claims (6)
1. A reactor core three-dimensional power rapid construction method based on a Monte Care critical computation one-step method is characterized by comprising the following steps:
step 1: determining an optimal population increment sequence according to a fission source error transfer model: obtaining a generic population increment sequence from a fission source error transfer model
Wherein: i represents the serial number of the calculation generation, and the values are 1, 2 and 3 …; m is(1)Is the initial particle number, m(i)Is the number of particles of the ith generation; c is a constant associated with the model by determining the constant c and the initial number of particles m(1)The optimal population increment sequence can be determined;
step 2: determining the optimal weight of each calculation generation on the overall calculation result according to the particle number increasing sequence, wherein the previous p generation fission source has a large error, and on the premise that the total particle number is fixed, the proportion of the previous p generation neutron statistical information to the total neutron statistical information is minimized to obtain an optimal weighting coefficient, so that the highest calculation efficiency is realized;
and step 3: performing Monte Care critical calculation and counting the three-dimensional power distribution of the reactor core: ensuring that the number of particles in each generation is equal to the optimal number-of-particles increasing sequence, no longer distinguishing the inactive generation from the active generation, counting the power distribution from the 1 st calculation generation, adopting the optimized weighting coefficient for each generation of weighting coefficients, and further constructing the three-dimensional power distribution of the reactor core
Wherein: i isCalculating the serial number of the generation; w is a(i)Is a weighting coefficient of the ith generation; m is(i)Is the number of particles of the ith generation; r (m)(i)) Is the power response function of the ith generation, Q(i)Is the power distribution of the ith generation, and N is the total number of computations.
2. The Monte Care critical computation one-step method-based reactor core three-dimensional power rapid construction method according to claim 1, wherein the inactive generation is not computed any more, and statistics of power distribution is performed from the first generation without considering fission source convergence diagnosis problems.
3. The Monte Care critical computation one-step method-based reactor core three-dimensional power rapid construction method according to claim 1, wherein the constant c is 1.0.
4. The Monte-Critical-computation-single-step-method-based reactor core three-dimensional power rapid construction method according to claim 1, wherein the initial particle number m is(1)Values of 50, 100 or 1000.
5. The reactor core three-dimensional power rapid construction method based on Monte Care Critical computation one-step method as claimed in claim 1, wherein the neutron statistical information of the previous p generations accounts for the portion of the total neutron statistical information
Wherein:
information quantity of previous p generations; scumIs the total information amount; n represents the total algebra of the simulation; w is a(i)Is a weighting coefficient of the ith generation; m is(i)Is the number of particles of the ith generation; c is a constant associated with the model.
6. The reactor core three-dimensional power rapid construction method based on the Monte-Critical computation one-step method as claimed in claim 1 or 4, wherein under the premise of fixed total particle number, the optimized weighting coefficient is obtained by minimizing the proportion of the previous p-generation neutron statistical information to the total neutron statistical information,
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