CN106128518A - A kind of method obtaining the few group cross-section of high-precision fast neutron reaction pile component - Google Patents

Abstract

A method for obtaining high-precision few-group cross-sections of fast neutron reactor components, by combining point-section data and multi-group data to calculate the fine-group cross-sections of fast neutron reactor components, calculating elastic scattering matrices online, and using fine Solve the neutron transport equation of the group cross-section to obtain the neutron flux density of the fine group, and then merge the energy group to obtain the few-group parameters of the fast reactor components; since this method directly uses the point cross-section data, the elasticity of the medium-mass nuclides The scattering resonance effect and the resonance interference effect of all nuclides are accurately considered; the neutron flux density of the real system is used in the process of energy group merging, and the merging result is more accurate; the method of the present invention calculates fewer fast neutron reactor components Compared with the reference value, the error of the group parameter is within 1% as a whole, which has a high precision.

Description

A Method for Obtaining Few Group Sections of Fast Neutron Reactor Components with High Accuracy

technical field

The invention relates to the fields of nuclear reactor core design and nuclear reactor physical calculation, in particular to a method for obtaining high-precision fast neutron reactor component few-group section.

Background technique

At present, the development of fast reactors is in the pre-design and experimental stage. In order to quickly and accurately solve the neutronics parameters of the core, the method based on the "two-step method" has become the main method for fast reactor engineering calculations. The so-called "two-step method", the first step is to carry out multi-group neutron transport calculations for various component materials in the core, obtain the neutron flux density distribution in the components, and then merge the few-group homogenization group cross-section; the second The first step is to solve the minority group neutron transport equation for the core with the homogenization parameters calculated in the previous step, and obtain physical quantities such as the effective multiplication factor of the core and the power distribution of the core.

In fast reactors, materials such as stainless steel and liquid metal sodium are relatively common and exist in large quantities. The elastic scattering cross sections of these nuclides have extremely strong elastic scattering resonance effects. Since the energy region where the elastic scattering resonance exists is exactly the region where the neutron flux density distribution of the fast neutron reactor is concentrated, this effect must be accurately considered. In addition, since the fission reaction is caused by fast neutrons, and the threshold energy reaction has a cross-section value of the same order as the elastic scattering cross-section, the threshold energy reaction should also be paid attention to; finally, not only heavy nuclides have resonance effects, but also nuclei with medium atomic mass Nuclides also have a resonance effect, and the resonance interference effect of multiple nuclides in the material should also be considered. The so-called resonance interference effect refers to the influence caused by the overlap of resonance peaks of two or more nuclides with resonance effect.

On the other hand, fast neutrons have a longer mean free path, and their local inhomogeneity effects are weak, and high enough accuracy can be obtained by considering the uniformity problem in the energy spectrum calculation of components. The so-called non-uniform effect means that due to the different cross-sectional values of different media such as fuel and coolant in the space, the distribution of neutron flux in different media has significant differences.

At present, the "two-step method" has been widely used in thermal reactors (such as pressurized water reactors) and has been recognized by the industry, and there are very mature calculation software. However, due to the great difference in neutron characteristics between fast reactors and thermal reactors, the methods used by these programs are not suitable for the calculation of fast neutron reactors. Therefore, it is necessary to develop a fast neutron reactor with high precision section method.

Contents of the invention

In order to overcome the problems existing in the above-mentioned prior art, the object of the present invention is to provide a method for obtaining a high-precision fast neutron reactor component few-group section. In order to obtain an accurate fast neutron reactor component few-group section, the method of the present invention Calculate the neutron flux density distribution in the module based on point section information, and then perform energy group merger to obtain the parameters of the minority group section. The data.

In order to achieve the above object, the present invention takes the following technical solutions:

A method for obtaining a high-precision fast neutron reactor component few-group section, comprising the steps of:

Step 1: For any fast reactor component to be calculated, read the information in the multi-group database, including the total cross section, inelastic scattering cross section matrix, neutron fission spectrum, and the number of neutrons released by each fission, using the formula (2) Calculate the macroscopic inelastic scattering matrix of the component, and use the formula (3) to calculate the dilution cross-section value of each nuclide;

In the formula:

——macroscopic inelastic scattering cross-section matrix of the gth group to the g'th group

N _i ——nucleon density of the first nuclide

——Inelastic scattering cross-section matrix of the first nuclide group g to g'

g g′—— fine group energy group identification number

In the formula:

- Dilution cross section of the nuclide

——Nuclide microscopic total cross-section

N _j —— nucleon density of the nuclide

Step 2: For any fast reactor component to be calculated, read the point section database information, including the interpolation table of total section, elastic scattering section, fission section and indistinguishable resonance region, and use formula (5) to calculate the fine group The microscopic average total cross section, microscopic average total elastic scattering cross section, and microscopic average fission cross section are interpolated using the dilution cross section of each nuclide calculated in step 1 according to the interpolation table provided by the database;

In the formula:

——the average cross section of group g, x is the type of reaction

Σ ^t ——The macroscopic total cross-section of the system

ΔE _g ——the energy group interval of the gth group

Step 3: From the microcosmic average total elastic scattering cross section of each nuclide provided in step 2, calculate the elastic scattering cross section matrix of each nuclide by using formula (8), calculate the macroscopic elastic scattering cross section matrix by using formula (10), and compare with step 1 The calculated macroscopic inelastic scattering cross-section matrix is summed, and the macroscopic scattering cross-section matrix is calculated by formula (11);

In the formula:

——the elastic scattering cross-section of the l-th order from the g-th group to the g'-th group

σ _s,g ——Elastic scattering cross section of group g

α——(A-1) ² /(A+1) ² , A is the atomic mass of the nuclide

μ _c —cosine value of scattering angle in centroid coordinate system

f(E,μ _c )——scattering probability function

P _l (μ _s )——Legendre polynomial of order μ _s

μ _s —cosine value of scattering angle in laboratory coordinate system

In the formula:

——The macroscopic elastic scattering cross-section matrix of the first-order group g to g'

——Elastic scattering cross-section matrix of the l-th nuclide from group g to group g'

In the formula:

——The macroscopic scattering cross-section matrix of the first-order group g to g'

Step 4: Obtain the microcosmic average total cross section, macroscopic scattering cross section matrix, and microcosmic average fission cross section of the fast reactor components from step 2 and step 3, use formula (12) to solve the neutron transport equation, and obtain the neutron flux of the fine group Quantity distribution, use formula (13) and (14) to carry out energy group merging of the section by using the obtained neutron flux distribution, obtain the low-group section value with high precision;

In the formula:

B - the critical curvature of the system

χ _g ——neutron fission spectrum of group g

υ - the number of neutrons released per fission

k _eff - the effective proliferation factor of the system

——The order neutron flux density distribution of group g

${\mathrm{<span\; class="notranslate">\; A</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}<span\; class="notranslate">\; =</span>{\mathrm{<span\; class="notranslate">\; b</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; -</span>\mathrm{<span\; class="notranslate">\; 1</span>}}<span\; class="notranslate">\; +</span>\frac{{\mathrm{<span\; class="notranslate">\; a</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}}{{\mathrm{<span\; class="notranslate">\; A</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}}$

${\mathrm{<span\; class="notranslate">\; a</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}<span\; class="notranslate">\; =</span>\frac{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}{\mathrm{<span\; class="notranslate">\; 2</span>}\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}\frac{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}{\mathrm{<span\; class="notranslate">\; 2</span>}<span\; class="notranslate">\; (</span>\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}<span\; class="notranslate">\; )</span><span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}{\mathrm{<span\; class="notranslate">\; B</span>}}^{\mathrm{<span\; class="notranslate">\; 2</span>}}$

${\mathrm{<span\; class="notranslate">\; b</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}<span\; class="notranslate">\; =</span>{\mathrm{<span\; class="notranslate">\; \&Sigma;</span>}}_{\mathrm{<span\; class="notranslate">\; t</span>}}<span\; class="notranslate">\; -</span>{\mathrm{<span\; class="notranslate">\; \&Sigma;</span>}}_{\mathrm{<span\; class="notranslate">\; the\; s</span>}}^{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}$

In the formula:

G - the identification number of the minority group energy group

——The thin group g belonging to the Gth minority group.

Compared with the prior art, the present invention has the following prominent advantages:

1. The process of making fine-group cross-sections of fast reactor components using the point-section database carefully considers the very strong elastic scattering resonance effect of medium-mass nuclides in the medium-to-high energy range and the resonance interference effect of all nuclides in the full-energy range .

2. When the point section is merged into fine group section, the real neutron flux density distribution of the problem is provided.

3. Calculate the fine-group neutron flux density distribution of the fast reactor components and merge the cross-sections accordingly, which not only saves calculation time, but also has sufficient accuracy.

Description of drawings

Figure 1 is a flow chart of computing the minority-group cross-section using point cross-sections.

Fig. 2 is a schematic diagram of iteratively solving nuclide dilution cross section.

Figure 3 is the macroscopic total section of 26 groups and their errors.

detailed description

Below in conjunction with accompanying drawing and specific embodiment the present invention is described in further detail:

In the present invention, by directly using the point section file, carefully considering the resonance effect existing in the fast reactor components, solving the neutron transport equation of the uniform problem, obtaining the real neutron flux distribution of the components, and then merging the energy groups to obtain Accurate minority group section parameter, as shown in Figure 1, this invention comprises the following steps:

Step 1: For any fast reactor component that needs to be calculated, read the information in the multi-group database, including the total cross section, inelastic scattering cross section, neutron fission spectrum, and the number of neutrons released by each fission, and then firstly calculate the temperature of the cross section Interpolation, where the temperature interpolation of the section follows the law, as shown in formula (1):

In the formula:

σ——section

x—response type, such as total cross section

T - temperature

Subscript 3——Interpolated temperature point

Subscript 1, 2——interpolation temperature point

The inelastic scattering matrix of each nuclide at a specific temperature obtained by interpolation is used to calculate the macroscopic inelastic scattering matrix of the component by formula (2).

In the formula:

——macroscopic inelastic scattering cross-section matrix of the gth group to the g'th group

N _i ——nucleon density of the first nuclide

——Inelastic scattering cross-section matrix of the first nuclide group g to g'

Subsequently, the dilution cross section value of each nuclide is calculated, and for the nuclide, its dilution cross section is defined by formula (3).

In the formula:

- Dilution cross section of the nuclide

——Nuclide microscopic total cross-section

N _j —— nucleon density of the nuclide

The dilution cross-section in formula (3) is determined by the total cross-section of other nuclides, and after the dilution cross-section of this nuclide is determined, the total cross-section value under this dilution cross-section will be obtained again. The whole process needs to be iteratively calculated. When the final An iteration is considered convergent when the total macroscopic cross-section of the problem hardly changes. The iterative process of dilution section calculation is shown in Figure 2:

1) Initialize the initial value of the dilution cross section of all nuclides, and perform interpolation calculation according to formula (4) according to the initial dilution cross section value, to obtain the total cross section value of each nuclide;

2) According to the new total cross-section value of each nuclide, use formula (3) to calculate the new dilution cross-section value of all nuclides;

3) Calculate the total cross-section value according to formula (4) from the newly generated dilution cross-section;

4) When the relative deviation of the total cross-section of the two calculations is less than 0.000001, the calculation is considered to be converged, otherwise the calculation continues.

Equation (4) gives the log-log interpolation law.

In the formula:

σ ^x —the cross-section of a certain reaction, such as the total cross-section

σ ⁰ ——Dilution cross section

Subscript 3——Interpolated point

Subscript 1, 2——interpolation point

Step 2: For any fast reactor component to be calculated, read the point section database information, including the interpolation table of the total section, elastic scattering section, fission section and indistinguishable resonance region; for the section located in the indistinguishable resonance region, use Step 1 calculates the dilution cross section of each nuclide obtained after convergence, and the total cross section, fission cross section, and elastic scattering cross section of each nuclide in the indistinguishable resonance region can be calculated by interpolation, so as to obtain the full energy range of all nuclides Point section information. Since the number of energy points in the cross-section of each nuclide is different, the point cross-section of each nuclide is calculated by linear interpolation in the full energy segment to obtain the value of the point cross-section under equal energy points. Based on the narrow resonance approximation, the multigroup cross section of each nuclide can be determined by formula (5). The so-called narrow resonance approximation means that the width of the resonance peak is narrow enough to be smaller than the width of energy lost after elastic scattering of neutrons.

In the formula:

——the average cross section of group g, x is the reaction type, such as the total cross section, etc.

Σ ^t ——The macroscopic total cross-section of the system

ΔE _g ——the energy group interval of the gth group

The average total cross-section, average total elastic scattering cross-section, and average fission cross-section of fine groups can be calculated by using formula (5);

Step 3: the microcosmic average total elastic scattering cross-section value of each nuclide provided by step 2, use the formula (8) to calculate the elastic scattering cross-section matrix of each nuclide, and use the formula (10) to calculate the macroscopic elastic scattering cross-section matrix of the problem, And add up with the macroscopic inelastic scattering cross-section matrix that step 1 calculates, utilize formula (11) to calculate macroscopical scattering cross-section matrix;

In general, after the elastic scattering reaction of neutrons with incident energy , the scattering cross section of outgoing neutrons with energy can be determined by the following formula:

In the formula:

σ _s (E→E′)——the elastic scattering cross section from energy point E to energy point E'

α——(A-1) ² /(A+1) ² , A is the atomic mass of the nuclide

μ _c —cosine value of scattering angle in centroid coordinate system

f(E,μ _c )——scattering probability function

Integrating the formula (6) over the energy group, the elastic scattering cross section from group to group can be obtained, as follows:

In the formula:

φ _g ——neutron flux density distribution of group g

Strictly solving formula (7) will consume a lot of time. On the basis of fine group and multi-group structure, an approximation can be introduced. It is considered that the neutron flux density and cross section are constant on a certain energy group, then formula (7) can be Simplifies to:

In the formula:

σ _s (g→g′)—the elastic scattering cross section from the g-th group to the g’-th group

σ _s,g ——Elastic scattering cross section of group g

When the elastic scattering cross section is expanded according to the Legendre polynomial, the scattering cross section of each order can be expressed as:

In the formula:

- the first order scattering cross section

P _l (μ _s )——Legendre polynomial of order μ _s

μ _s —cosine value of scattering angle in laboratory coordinate system

Finally, according to the formula (9), the elastic scattering cross-section matrices of all nuclides can be obtained, and the macroscopic elastic scattering cross-section matrix of the system can be calculated by using the formula (10). The elastic scattering cross-section matrix, the macroscopic scattering cross-section matrix of the system can be obtained by using the formula (11).

In the formula:

——The macroscopic elastic scattering cross-section matrix of the first-order group g to g'

——Elastic scattering cross-section matrix of the l-th nuclide from group g to group g'

In the formula:

——The macroscopic scattering cross-section matrix of the first-order group g to g'

Step 4: Obtain the microscopic average total section, macroscopic scattering cross section matrix, and microscopic average fission section of the fast reactor components from step 2 and step 3, and use formula (12) to solve the multi-group neutron transport equation for the calculated fast reactor components The fine-group and multi-group neutron flux density distribution of this problem can be obtained:

In the formula:

B - the critical curvature of the system

χ _g ——neutron fission spectrum of group g

υ - the number of neutrons released per fission

k _eff - the effective proliferation factor of the system

——The order neutron flux density distribution of group g

${\mathrm{<span\; class="notranslate">\; A</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}<span\; class="notranslate">\; =</span>{\mathrm{<span\; class="notranslate">\; b</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; -</span>\mathrm{<span\; class="notranslate">\; 1</span>}}<span\; class="notranslate">\; +</span>\frac{{\mathrm{<span\; class="notranslate">\; a</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}}{{\mathrm{<span\; class="notranslate">\; A</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}}$

${\mathrm{<span\; class="notranslate">\; a</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}<span\; class="notranslate">\; =</span>\frac{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}{\mathrm{<span\; class="notranslate">\; 2</span>}\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}\frac{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}{\mathrm{<span\; class="notranslate">\; 2</span>}<span\; class="notranslate">\; (</span>\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}<span\; class="notranslate">\; )</span><span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}{\mathrm{<span\; class="notranslate">\; B</span>}}^{\mathrm{<span\; class="notranslate">\; 2</span>}}$

${\mathrm{<span\; class="notranslate">\; b</span>}}_{\mathrm{<span\; class="notranslate">\; no</span>}}<span\; class="notranslate">\; =</span>{\mathrm{<span\; class="notranslate">\; \&Sigma;</span>}}_{\mathrm{<span\; class="notranslate">\; t</span>}}<span\; class="notranslate">\; -</span>{\mathrm{<span\; class="notranslate">\; \&Sigma;</span>}}_{\mathrm{<span\; class="notranslate">\; the\; s</span>}}^{\mathrm{<span\; class="notranslate">\; no</span>}<span\; class="notranslate">\; +</span>\mathrm{<span\; class="notranslate">\; 1</span>}}$

After the neutron flux density distribution is obtained, the energy group merging is performed to finally obtain the cross-sectional parameters of the few-group component of the fast reactor component. Using the neutron flux density as the weight to merge the calculation of the minority group cross-section parameters can be obtained by formulas (13) and (14). Among them, the formula (13) is used to merge the total cross section and the neutron generation cross section, and the formula (14) is used to merge the order scattering matrix.

In the formula:

G - the identification number of the minority group energy group

——a thin group g belonging to the Gth minority group

Fig. 3 shows the macro total section of 26 groups of fast reactor components calculated by the present invention and the cross section of 26 groups calculated by Monte Carlo method and the error between the two. Preliminary calculation results show that, compared with the reference solution, the relative error of the cross section of almost all energy groups is within 1%, and the error is controlled within 3% for individual energy groups with lower energy. Since the neutron flux density distribution at lower energies is also lower, this error is acceptable. Various neutronics calculations of the fast reactor core, such as steady-state calculations, burnup calculations, and transient calculations, can be performed using accurate minority-group cross-sections. The invention can obtain very high precision, and at the same time, the calculation efficiency is far higher than that of the Monte Carlo method, and can be applied to actual engineering calculations.

Claims (1)

1. the method obtaining the few group cross-section of high-precision fast neutron reaction pile component, it is characterised in that: comprise the steps:

Step 1: for the arbitrary fast pile component of required calculating, reads information in Multi-group data storehouse, including total cross section, non-resilient dissipates Penetrate cross section matrix, neutron fission spectrum, every time fission release neutron population, utilize macroscopical inelastic scattering of formula (2) computation module Matrix, utilizes formula (3) to calculate the dilution cross section value of each nucleic；

In formula:

G group is to macroscopical inelastic scattering cross section matrix of g ' group

N_iThe nucleon density of the nucleic

The nucleic g group is to the inelastic scattering cross section matrix of g ' group

G g' thin group energy group identification number

In formula:

The dilution cross section of nucleic

Nucleic microcosmic total cross section

N_jThe nucleon density of nucleic

Step 2: for the arbitrary fast pile component of required calculating, reads some library of cross section information, including total cross section, elastic scattering Cross section, fission cross section and the interpolation table of resonance region can not be differentiated, utilize formula (5) calculate thin group the average total cross section of microcosmic, Microcosmic average proof resilience scattering section, microcosmic Average Fission cross section, utilize the dilution of calculated each nucleic in step 1 to cut The interpolation table that face provides according to data base carries out interpolation；

In formula:

The averga cross section of g group, x is response type

Σ^tThe volumic total cross-section of system

ΔE_gG group's can group interval

Step 3: the microcosmic average proof resilience scattering section of each nucleic provided by step 2, utilizes formula (8) to calculate each core The elastic scattering cross-section matrix of element, utilizes formula (10) to calculate macroscopic view elastic scattering cross-section matrix, and calculated with step 1 Macroscopic view inelastic scattering cross section matrix add and, utilize formula (11) to calculate macroscopic scattering cross section matrix；

In formula:

L rank are scattered to the elastic scattering cross-section σ of g ' group by g group_s,gThe elasticity of g group dissipates Penetrate cross section

α——(A-1)²/(A+1)², A is the atomic mass of nucleic

μ_cAngle of scattering cosine value under geocentric coordinate system

f(E,μ_c) probability of scattering function

P_l(μ_s) about μ_sRank Legnedre polynomial

μ_sAngle of scattering cosine value under laboratory coordinate

In formula:

L rank g group is to macroscopical elastic scattering cross-section matrix of g ' group

L rank the nucleic g group is to the elastic scattering cross-section matrix of g ' group

In formula:

L rank g group is to the macroscopic scattering cross section matrix of g ' group

Step 4: obtained the average total cross section of microcosmic of fast pile component by step 2 and step 3, macroscopic scattering cross section matrix, microcosmic are put down All fission cross section, utilizes formula (12) to carry out solving of neutron-transport equation, it is thus achieved that the Neutron flux distribution of thin group, utilization is tried to achieve Neutron flux distribution use that formula (13) and (14) carries out cross section can group's merger, it is thus achieved that there is high-precision few group cross-section Value；

${\mathrm{iB\&phi;}}_1^g+{\mathrm{\&Sigma;}}_t^g{\mathrm{\&phi;}}_0^g=\underset{g^{\&prime;}}{\&Sigma;}{\mathrm{\&Sigma;}}_{s,0}^{g^{\&prime;}\&RightArrow;g}{\mathrm{\&phi;}}_0^{g^{\&prime;}}+\frac{{\mathrm{\&chi;}}_g}{k_{eff}}\underset{g^{\&prime;}}{\&Sigma;}{\mathrm{\&upsi;\&Sigma;}}_f^{g^{\&prime;}}{\mathrm{\&phi;}}_0^{g^{\&prime;}}$

${\mathrm{\&phi;}}_l^g=-\frac{1}{(2l+1)A_l^g}{\mathrm{iB\&phi;}}_{l-1}^g,l=2,3,...,N$

In formula:

The critical buckling of B system

χ_gThe neutron fission spectrum of g group

υ fissions release neutron population every time

k_effThe Effective multiplication factor of system

G order of a group netron-flux density is distributed

$A_n=b_{n-1}+\frac{a_n}{A_{n+1}}$

$a_n=\frac{n+1}{2n+1}\frac{n+1}{2(n+1)+1}{B}^2$

$b_n={\mathrm{\&Sigma;}}_t-{\mathrm{\&Sigma;}}_s^{n+1}$

In formula:

The few group energy group identification number of G

Belong to the thin group g of the few group of G.