CN117150829A - Method for generating stack-used few-group cross-section energy group structure for neutron analysis of intermediate energy spectrum - Google Patents
Method for generating stack-used few-group cross-section energy group structure for neutron analysis of intermediate energy spectrum Download PDFInfo
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Abstract
A method for generating a few-group cross-section energy group structure for a reactor for neutron analysis of an intermediate energy spectrum, which needs to merge a plurality of groups of cross sections adopted by component calculation into a few groups of cross sections before core neutron calculation is carried out. Therefore, the method of the invention firstly establishes an equivalent reactor core model when the assembly is calculated, then calculates neutron production rate of a corresponding multi-group structure in the model, normalizes to obtain neutron production probability of Xiang Yingneng groups, integrates the obtained neutron production probability according to energy to obtain neutron production share of a low energy region, judges whether the reactor core energy spectrum belongs to an intermediate energy spectrum or not, and divides the energy groups by taking the neutron production probability as a standard when the energy spectrum is the intermediate energy spectrum to obtain a final few-group energy group structure. The purpose of this is to accurately capture the energy spectrum characteristics of the core when the few cross sections are generated, and to improve the accuracy of the merging of the few cross sections.
Description
Technical Field
The invention relates to the field of nuclear reactor core physical computation, in particular to a method for generating a few-group cross-section energy group structure for a reactor for neutron analysis of an intermediate energy spectrum.
Background
In recent years, a great deal of advanced reactor technology breaks through, and a series of innovative conceptual schemes including small modular reactors, special micro-reactors and the like are proposed, but some problems are brought about. First, core neutron spectrum characteristics have changed significantly due to core miniaturization and the application of solid moderators. On the one hand, the average neutron energy of fission induced in the reactor is far smaller than that of a typical fast reactor but far larger than that of a typical thermal neutron reactor; on the other hand, the energy is variable in the energy range from 1eV to 1keV along with the change of the reactor core design. Therefore, neutron calculation methods and calculation programs suitable for conventional pressurized water reactors and fast reactors have failed to meet the calculation requirements caused by such changes. The applicability of intermediate energy spectrum becomes the most critical technical problem to be solved in the field of core physical calculation and numerical analysis.
In order to obtain few-cluster cross sections with better energy spectrum applicability, a new generation of reactor component programs has expanded the component multi-cluster calculation to thousands of clusters or more. However, the energy group structure cannot be directly used for neutron analysis in the reactor core at present, and needs to be first integrated into a few group structure. In the merging process, the information of the fine energy spectrum during the component calculation is inevitably lost, and the reactor core becomes extremely sensitive to specific energy spectrum characteristics. Experience shows that the error brought by different energy group structures can reach thousands of pcm, which exceeds the allowable error range of the reactor core design. On the other hand, due to uncertainty of the new concept reactor design, specific neutron spectrum characteristics of the reactor core cannot be predicted in advance, and the empirically-based and determined few-group energy group structure cannot meet the requirement of the new concept reactor design for variability. Therefore, how to obtain the few-group energy group structure for neutron analysis in the reactor core, which can meet the changeable intermediate energy spectrum applicability, reduces errors caused by energy group merging in the process of calculating from the assembly to the reactor core, and has important significance and value.
Disclosure of Invention
In order to overcome the problems in the prior art, the invention aims to provide a generation method of a few-group cross-section energy group structure for reactor for neutron analysis of intermediate energy spectrum, which is used for adaptively generating a matched few-group structure according to specific reactor core characteristics, so that the loss of reactor core energy spectrum characteristics in the energy group merging process is avoided, and the accuracy of the few-group cross section for neutron analysis of the reactor core is ensured.
In order to achieve the above purpose, the invention adopts the following technical scheme:
a method for generating a few-group cross-section energy group structure for a reactor for neutron analysis in an intermediate energy spectrum, which uses an equivalent reactor core model to calculate and obtain a plurality of groups of neutron production rates, then judges energy spectrum characteristics by neutron production shares in a low energy region, calculates neutron production probability, and divides energy groups according to the sizes of the energy groups, comprises the following steps: step 1: performing volume equivalence on the radial geometry of the reactor core according to the material area, and neglecting the axial geometry to obtain an equivalent reactor core model;
step 2: calculating the equivalent reactor core model obtained in the step 1 to obtain a plurality of groups of neutron production rates, and normalizing to obtain neutron production probability;
(1)
wherein:
-neutron production probability of group g;
-neutron production rate in multiple clusters in the unit of；
-nuclear density of each nuclide i in units of；
Fissile section of group g of nuclide i in units of；
-neutron flux density of group g, in units of；
-the average number of fission neutrons generated by each fission of group g of nuclide i;
step 3: integrating the neutron generation probability obtained in the step 2 according to energy to obtain a neutron generation share in a low energy region;
(2)
wherein:
-neutron production fraction in the low energy region;
-lower energy limit of low energy region;
-an upper energy limit of the low energy region;
step 4: judging whether the energy spectrum belongs to the intermediate energy spectrum according to the neutron generation share of the low energy region obtained in the step 3;
when 1% < X <50%, namely the neutron generation proportion of the low energy region is 1% -50%, the energy spectrum is judged to belong to the middle energy spectrum, and therefore the multi-group structure is divided according to the neutron generation probability;
step 5: the energy groups of different energy spectrum characteristics and different energy segments are adaptively divided according to the size of neutron generation probability;
step 6: obtaining the final energy group structure of the few groups.
In step 5, energy groups of different energy spectrum characteristics and different energy segments are adaptively divided according to the size of neutron generation probability, and the method specifically comprises the following steps:
dividing the high-energy region, if 1% < X <30%, equally dividing neutron generation probability into 10 sections according to the size, dividing energy groups in the same section into a few energy groups, wherein the upper limit of the number of the energy groups in one few energy groups is 100, and the lower limit of the number of the energy groups in one few energy groups is 60; if 30% < X <50%, equally dividing neutron production rate into 2 sections according to the size, dividing the energy groups in the same section into one less energy group, wherein the upper limit of the number of the energy groups in one less energy group is 200, and the lower limit of the number of the energy groups in one less energy group is 60;
dividing a low-energy region, and dividing neutron generation probability according to 1:15 is divided into 2 sections, the energy groups in the same section are divided into a few energy groups, and the upper limit of the number of the energy groups in one few energy groups is 200, and the lower limit is 15;
for the thermal neutron energy region, a fixed division mode is adopted, and the division mode is as follows:
the numbers of the upper and lower boundaries of the energy group are 1915-1926, 1927-1935, 1936-1941, 1942-1951, 1952-1953, 1954-1956, 1957-1960, 1961-1964, 1965-1968, respectively.
Compared with the prior art, the invention has the following outstanding advantages:
1. in the method, when the energy group is calculated from the multi-group calculation of the components to the few-group calculation of the reactor core, the energy group structure is adaptively divided according to the size of neutron generation probability instead of the traditional fixed energy group structure, so that the loss of the energy spectrum characteristics of the reactor core in the energy group merging process is avoided, and the accuracy of the few-group section for neutron analysis of the reactor core is ensured;
2. in the method, neutron generation probability is provided, and the energy group structure is divided according to the size of the neutron generation probability, so that contribution shares of different energy groups in core calculation can be accurately reflected, and further, energy spectrum differences of different cores and the influence of the energy spectrum differences on the neutron analysis are more accurately reflected;
3. in the method, an equivalent reactor core model is adopted to replace an actual reactor core model, so that neutron generation probability distribution which reflects main energy spectrum characteristics of the reactor core can be obtained efficiently, and a foundation is laid for self-adaptive minority structure generation.
Drawings
FIG. 1 is a flow chart of the method of the present invention.
Fig. 2 is a schematic diagram of core volume equivalence.
FIG. 3 is a schematic diagram of a neutron production probability division energy group.
Detailed Description
The invention is described in further detail below with reference to the attached drawings and detailed description:
for the reactor core shown on the left of fig. 2, the equivalent calculation is performed on the core by using the invention to obtain a few-group cross-section energy group structure suitable for the core, and as shown in fig. 1, the steps are as follows:
step 1: performing volume equivalence on the radial geometry of the reactor core at the left side of FIG. 2 according to a material region, and neglecting the axial geometry to obtain a reactor core model of the equivalent reactor at the right side of FIG. 2, wherein the model is circular;
step 2: calculating the equivalent reactor core model obtained in the step 1 to obtain 1968 group neutron production probability;
(1)
wherein:
-neutron production probability of group g;
-neutron production rate in multiple clusters in the unit of；
-nuclear density of each nuclide i in units of；
Fissile section of group g of nuclide i in units of；
-neutron flux density of group g, in units of；
-the average number of fission neutrons generated by each fission of group g of nuclide i;
step 3: the neutron production probability obtained in the step 2 is 10 according to the energy -5 Integrating eV-100 eV to obtain neutron generation share X=49.4% in the low energy region;
(2)
wherein:
-neutron production fraction in the low energy region;
-lower energy limit of low energy region;
-an upper energy limit of the low energy region;
step 4: judging that the energy spectrum of the reactor core belongs to an intermediate energy spectrum (1% < X < 50%) according to the neutron generation share of the low energy region obtained in the step 3;
step 5: the energy groups of different energy spectrum characteristics and different energy segments are divided according to neutron generation probability:
dividing 1-1479 groups (100 eV-19.6 MeV), equally dividing neutron generation probability into 2 sections according to the size as 30% < X <50%, and dividing the energy groups in the same section into one few energy groups, wherein an example is shown in FIG. 3;
1480-1915 groups (0.245 eV-100 eV) are divided, and neutron generation probability is 1:15 is divided into 2 segments, and the energy groups in the same segment are divided into a few energy groups.
For groups 1915-1968 (10) -5 eV-0.625 eV), adopting a fixed division mode:
the numbers of the upper and lower boundaries of the energy group are 1915-1926, 1927-1935, 1936-1941, 1942-1951, 1952-1953, 1954-1956, 1957-1960, 1961-1964, 1965-1968, respectively.
Step 6: obtaining the final energy group structure of the few groups.
In the definite theory program, the minority group energy group structure is utilized to calculate the minority group section of the reactor assembly to obtain the corresponding minority group section, and finally the reactor core calculation is carried out to obtain the reactor core k eff The results are shown in Table 1: the results show that the few-group energy group structure obtained by using the method can remarkably reduce calculation deviation.
TABLE 1 comparison of the results of the present invention with the existing 33-group energy group structure calculation
The invention can meet the calculation requirement of the middle energy spectrum reactor core, obtain the neutron production probability distribution of the reactor core by using the equivalent reactor core model, efficiently obtain the middle energy spectrum characteristic of the reactor core, divide the energy groups by taking the size of the neutron production probability as the basis, accurately consider the contribution share of different energy groups in the reactor core calculation, avoid the loss of the middle energy spectrum characteristic when merging from multiple groups of components to few energy groups of the reactor core under the traditional fixed few energy groups structure, and provide a new idea for the physical calculation and analysis of the middle energy spectrum reactor core.
Claims (2)
1. A method for generating a few-group cross-section energy group structure for stacking in intermediate energy spectrum neutron analysis is characterized by comprising the following steps of: the method comprises the following steps:
step 1: performing volume equivalence on the radial geometry of the reactor core according to the material area, and neglecting the axial geometry to obtain an equivalent reactor core model;
step 2: calculating the equivalent reactor core model obtained in the step 1 to obtain a plurality of groups of neutron production rates, and normalizing to obtain neutron production probability;
(1)
wherein:
-neutron production probability of group g;
-neutron production rate in ∈k->；
-nuclear density of each nuclide i in +.>；
-fissile section of group g of nuclide i in +.>；
-neutron flux density of group g in +.>；
-the average number of fission neutrons generated by each fission of group g of nuclide i;
step 3: integrating the neutron generation probability obtained in the step 2 according to energy to obtain a neutron generation share in a low energy region;
(2)
wherein:
-neutron production fraction in the low energy region;
-lower energy limit of low energy region;
-an upper energy limit of the low energy region;
step 4: judging whether the energy spectrum belongs to the intermediate energy spectrum according to the neutron generation share of the low energy region obtained in the step 3;
when 1% < X <50%, namely the neutron generation proportion of the low energy region is 1% -50%, the energy spectrum is judged to belong to the middle energy spectrum, and therefore the multi-group structure is divided according to the neutron generation probability;
step 5: the energy groups of different energy spectrum characteristics and different energy segments are adaptively divided according to the size of neutron generation probability;
step 6: obtaining the final energy group structure of the few groups.
2. The method for generating a stack-used minority group cross-sectional energy group structure for intermediate spectroscopy neutron analysis according to claim 1, wherein: in step 5, energy groups of different energy spectrum characteristics and different energy segments are adaptively divided according to the size of neutron generation probability, and the method specifically comprises the following steps:
dividing the high-energy region, if 1% < X <30%, equally dividing neutron generation probability into 10 sections according to the size, dividing energy groups in the same section into a few energy groups, wherein the upper limit of the number of the energy groups in one few energy groups is 100, and the lower limit of the number of the energy groups in one few energy groups is 60; if 30% < X <50%, equally dividing neutron production rate into 2 sections according to the size, dividing the energy groups in the same section into one less energy group, wherein the upper limit of the number of the energy groups in one less energy group is 200, and the lower limit of the number of the energy groups in one less energy group is 60;
dividing a low-energy region, and dividing neutron generation probability according to 1:15 is divided into 2 sections, the energy groups in the same section are divided into a few energy groups, and the upper limit of the number of the energy groups in one few energy groups is 200, and the lower limit is 15;
for the thermal neutron energy region, a fixed division mode is adopted, and the division mode is as follows:
the numbers of the upper and lower boundaries of the energy group are 1915-1926, 1927-1935, 1936-1941, 1942-1951, 1952-1953, 1954-1956, 1957-1960, 1961-1964, 1965-1968, respectively.
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