JPH0772282A - Method and device for estimating reactor core performance - Google Patents

Abstract

PURPOSE:To improve accuracy by combining tabulated reactivity coefficients and the microscopic cross sectional areas based on the atom density and lattice calculation of Pu isotope and Am isotope which are calculated in actual power history conditions. CONSTITUTION:Obtained is the variation of nuclear constants caused by the difference in atom density of Pu isotope and Am isotope in a small region which is produced from the difference between the power history data for each node region in core and the power history data assumed when a burning equation of nuclides in the small region is solved. Then, by correcting the nuclear constant file, the count value from a neutron flux measuring system 41 and a state measuring system 48 in the reactor 46 are taken in a data sampling device 42 as inputs to calculate the power density distribution in the core 47. The results evaluated with a process computer 43 after being transmitted periodically by the data sampling device 42 to the process computer 43 are transmitted to an output device 45.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a reactor core performance estimation method and apparatus effective for use in reactor core design and core operation management.

[0002]

2. Description of the Related Art A method for estimating core performance of a nuclear reactor is shown in FIG.
Will be described with reference to. In FIG. 9, a core 21 for performance evaluation in a nuclear reactor is spatially divided into 22 small regions, and an equation representing the transport phenomenon of neutrons in the small regions is assumed by assuming boundary conditions at the outer boundaries of the small regions. To solve, set 23 the presence or absence of control rods, moderator density, fuel and moderator temperatures, and boundary conditions.

Then, a lattice calculation 24 is performed to create a nuclear constant table 25 for the small region, and the nuclear constant table 25 is used to perform fitting by burnup and moderator density, and a nuclear constant file used in the core performance calculation. After creating 26, core performance calculation 27 is performed.

Incidentally, reference numeral 28 in FIG. 9 includes reference numerals 26 and 27, which is a method for obtaining the core performance of a nuclear reactor, and 29 indicates a burnup, power density and moderator density file.

Here, the conventional example will be described in more detail as follows. In the radial direction of the reactor core 21 to be evaluated for the performance of the nuclear reactor, the area is made small 22 in units of fuel assemblies, and boundary conditions are assumed at the boundary of the fuel assemblies in a state where the fuel assemblies continue infinitely in the axial direction. The transport equation representing the neutron transport in the fuel assembly is solved for the combination of parameters such as the presence or absence of control rods, the density of the moderator, the temperature of the fuel and moderator (reference numeral 23).

At this time, the combustion equation is solved assuming the output history of the fuel assembly, the temporal change of the atomic number density of the nuclide, the nuclear constant table of the fuel assembly are calculated in advance, and the nuclear constant table 25 is prepared. (Hereafter, this calculation is called grid calculation).

After that, a nuclear constant file 26 used for core performance calculation prepared by using the nuclear constant table 25 of the fuel assembly
A method of determining the core performance of the reactor by solving the equation representing the neutron transport phenomenon of the entire reactor core and the equation representing the thermohydraulic behavior of the moderator or the empirical formula (reference numeral 27) based on Is taken.

Here, a calculation method called a node method, which is performed as a method for estimating the neutron flux distribution in the core, which is necessary when the equation representing the neutron transport phenomenon of the entire reactor is solved to evaluate the core performance Will be described.

In the node method, the core of a nuclear reactor is divided into fuel assembly units in the radial direction and axially divided into 24 lattice points called nodes, and the energy of neutrons is further divided into a plurality of groups, and each space is divided into a plurality of groups. Consider the neutron flux that represents each energy group at the lattice point. An unknown vector consisting of all neutron flux values representing all lattice points and all energy groups If so, the problem of obtaining the neutron flux distribution in the core is
It becomes a problem to solve simultaneous linear equations such as equation (1).

[0010]

[Equation 1] Here, M and F are matrices having dimensions equal to the number obtained by multiplying the number of all spatial grid points by the number of all energy groups, and k is an eigenvalue of the matrix (M ⁻¹ · F).

[0011]

The matrices M and F included in the simultaneous linear homogeneous equations (1) used in the conventional core performance estimation method described above are composed of nuclear constants evaluated by lattice calculation.

That is, matrix elements M ^g _{i, j, k} and F ^g
_{i, j, k} are expressed as the following equations.

[0013]

[Equation 2]

In the lattice calculation, the nuclear constants (Σ ^g _m and K ^m _∞ ) are normally decelerated with the burnup assuming the rated power density and output history without a fixed period (hereinafter referred to as constant output history) for each fuel type. Create as a function of material density.

However, in an actual nuclear reactor, not only is the operation stopped for each operating cycle to perform a periodic inspection (hereinafter referred to as “constant inspection”), but the output level of the nuclear reactor also fluctuates during operation. The power density greatly changes with the burnup at each point (hereinafter referred to as the actual power history).

The conventional core calculation is performed based on the node constant obtained by fitting the burnup and the history moderator density of each node from the nuclear constant evaluated by the grid calculation under the constant output history condition.

In the fuel, there are nuclides in which the atomic number density greatly changes depending on the output history, and the difference in the atomic number density generated between the constant output history and the actual output history has a great influence on the nuclear constant. There are nuclides that cannot be ignored in terms of core design and operation management.

In the conventional core calculation, Xe-135, Sm
For the two nuclides of -149, the atomic number density corresponding to the actual output history is calculated, and the magnitude of the influence of the difference with the atomic number density evaluated by the lattice calculation under the constant output history condition on the nuclear constant is evaluated. Corrects the constant.

Regarding the calculation method of the atomic number density of Xe-135 and Sm-149, for example, James J. Duderstat & Louis.
J. Hamilton "Nuclear Reactor Analysis" (1976) John
Wiley & Sons, Inc Chapter 15

However, in the fuel, Xe-135,
In addition to Sm-149, if the power density changes greatly during operation, or if there is a regular inspection during the operation cycle, the atomic number density will be significantly different from that under constant output history conditions, and nuclides that have a large effect on the nuclear constant Although they exist, no correction of the node constants has been made for these nuclides.

Particularly, in the case of MOX fuel in which a large amount of Pu isotope is contained in the initial fuel composition, Pu is contained during the reactor shutdown period.
It has been known that the reactivity change due to the β-decay of -241 and the production of Am-241 is large and must be taken into consideration in the core operation management. However, in the conventional core calculation, the atomic number density of these nuclides is known. There is a problem that cannot be considered because the change due to the output history of is not calculated.

In addition, some of the fission products include Gd-155.
There is a large thermal neutron absorption cross-section as shown in Fig. 3, and the reactivity effect of the amount produced by the decay of the preceding nucleus during the outage cannot be ignored for core management. Since the reactivity change of the product is not calculated,
There are issues that cannot be considered.

The present invention has been made to solve the above-mentioned problems, and Pu generated between the actual output history and the constant output history is included in the kernel constants included in the matrix element of the equation (1) handled in the node calculation.
Evaluating the difference in atomic number density between isotopes and Am isotopes,
The object of the present invention is to provide a highly accurate method for promoting core performance of a nuclear reactor by correcting the magnitude of the influence on the nuclear constant to a node constant.

Further, the reactivity effect caused by the difference between the actual output history and the constant output history condition is corrected to the nuclear constant, so that it is also high for the core which is operated significantly different from the output history assumed in the grid calculation. An object of the present invention is to provide a reactor core performance estimation device capable of maintaining accuracy.

[0025]

A method of estimating core performance of a reactor according to the present invention spatially divides a reactor core for performance evaluation of a reactor into small regions, and assumes boundary conditions at outer boundaries of the small regions. Then, the equation representing the transport phenomenon of neutrons in the small region is solved for the combination of parameters such as the presence of the control rod, the density of the moderator, the temperature of the fuel and the moderator, and the nuclear constant of the small region is calculated in advance. And create a nuclear constant table, and use this nuclear constant table to solve the node equation representing the neutron transport phenomenon of the entire reactor core and the equation or empirical formula representing the thermo-hydraulic behavior of fuel and moderator. In the reactor core performance estimation method, in solving the output history of the small region and the combustion equation of the nuclide in the small region obtained by solving the node equation representing the transport phenomenon of neutrons in the entire core The nuclear constants described in the nuclear constant table are obtained by obtaining the amount of change in the nuclear constants due to the difference in the atomic number densities of the Pu and Am isotopes in the small region resulting from the difference between the determined output history. And a nuclear constant table described in the nuclear constant table, the nuclear constant is corrected by the first means, and the correction is used to transport neutrons in the entire reactor core. From the third means for solving the node equation representing the phenomenon to obtain the power density distribution in the reactor and the fourth means for taking in the power density distribution and calculating at least the density of the moderator and the fuel temperature in the reactor. It is characterized by

Further, the reactor core performance estimating apparatus according to the present invention comprises a state quantity measuring system and a neutron flux measuring system installed in the reactor, a data sampling apparatus for inputting measured values of these measuring systems, It is characterized by comprising a process computer connected to the data sampling device, and an external storage device and an input / output device connected to the process computer.

Further, the reactor core performance estimating apparatus according to the present invention is a computer for inputting the operation plan of the reactor as input, data for describing the shape of the reactor core in advance connected to this computer, and cores of the materials constituting the reactor core. It was equipped with an external storage device that describes the nuclear constants that represent the characteristics, the thermal-hydraulic constants of the core, and other necessary physical property values, and an input / output device that is connected to the computer and that inputs the operation plan of the future nuclear reactor. Characterize.

[0028]

The energy first group combustion equation for the nuclide existing in the fuel assembly is expressed by the following equation.

[0029]

[Equation 3]

However, the conversion from the burnup E to the time t in the equation (3) is performed by the following equation.

[Equation 4]

The atomic number densities of the nuclide i obtained by solving the equation (3) for the actual output history and the constant output history assumed in the lattice calculation are N ^a _i (t) and N ^o _i (t), respectively. Then, the macroscopic sectional area Σ ^g of the g group and the correction amount ΔΣ ^g to the infinite multiplication factor K _∞ caused by the difference in the output history can be approximated by the following equation.

[0032]

[Equation 5]

[0033]

[Equation 6]

That is, in the core performance estimation method according to the present invention, the matrix element M ^g in the simultaneous linear homogeneous equations (1) is
_{i, j, k} and F ^g _{i, j, k} are expressed by the following equations.

[0035]

[Equation 7]

For the method of evaluating the atomic number density for each node in the core performance calculation, see, for example, Lindermeiter et al, A.
dvanced Methodology for LWR Core Design, Vol1P101
-107, Proceedings of the National Reactor Physics A
NS, Jackson Hole, Wyoming, September 18-22,1988, but in the method in this paper, the combustion equation is treated by two groups of energies, and the nuclide component of the nuclear constant is calculated by the following equation.

[0037]

[Equation 8]

However, the method described in this paper has a problem that the memory of the computer becomes enormous because the microscopic cross-sectional area library of the main nuclide and the atomic number density file are used.

On the other hand, in the present invention, the correction is made by using the difference in the atomic number density of the specific nuclide caused by the difference in the output history. The nuclear constants to be corrected are stored in a file for each fuel type, so the memory requirement is smaller than the method described in the above paper, and in the combustion equation, as in equation (3), Since energy is handled as one group, there is a feature that the calculation time is short.

In the prior art, the nuclear constant used in the core calculation is used by fitting the nuclear constant evaluated in advance by lattice calculation with the burnup and the moderator density. On the other hand, in the present invention, the atomic number density of the Pu isotope and the Am isotope can be made to correspond to the actual output history when creating the nuclear constant for each node, so that it is higher than that of the conventional technique. The core performance can be estimated with accuracy.

However, since the data used for correcting the nuclear constants is created based on the lattice calculation, when the present invention is carried out, a lattice calculation is newly added in addition to the lattice calculation performed in the conventional technique. No need.

The in-core three-dimensional output and moderator density distribution used when preparing the nuclear constant file are not evaluated by the core calculation, but are created based on plant data obtained through a process computer or the like. You can also use it.

[0043]

EXAMPLE A first example of the reactor core performance estimation method according to the present invention will be described. The first embodiment corresponds to claim 1, and the gist thereof will be described with reference to FIG. In the block of the flow chart in FIG. 1, the part marked with * in the upper right part is a part newly added in the present invention, and the unmarked part is the part of the conventional example.

In FIG. 1, the core 1 for performance evaluation to the nuclear constant table 5 of the small region are the same as the conventional example, and the nuclear constant file 7 and the core performance calculation 12 used in the core performance calculation are the same as those of the conventional example. It is the same. The part of the present invention different from the conventional example is Pu.
Combustion coefficients P _i , D _i for isotopes and Am isotopes,
This means that the reactivity coefficient ∂K _∞ / ∂N _i , the tabularization 6 of the microscopic cross-sectional area σ, the portions indicated by reference numerals 8 to 11 and the portion of the atomic number density file 14 are added.

Here, as a first means, the output history obtained by solving the node equation representing the neutron transport phenomenon of the whole core, that is, the output history data of each node area in the core and the nuclide in the small area Due to the difference in atomic number density of Pu and Am isotopes in the small region resulting from the difference between the output history assumed when solving the combustion equation of Eq. The amount of change in the nuclear constant is calculated, and the nuclear constant shown in Table 5 is corrected by the correction 10 of the nuclear constant file used in the core performance calculation.

As a second means, the nuclear constants shown in Table 5 are taken in (see reference numeral 7), the nuclear constant file used in the core performance calculation is corrected 10, and the nuclear constant file is used for correction. Create a coefficient table of the node equation that expresses the neutron transport phenomenon used in the core performance calculation of the entire core (see reference numeral 11).

As a third means, coefficients are taken from the coefficient table shown by reference numeral 11 and the equation representing the transport phenomenon of neutrons handled in the core performance calculation 12 is solved to obtain the power density distribution in the reactor (reference numerals 11 and 12). reference).

As a fourth means, the power density distribution is taken in and the moderator density and fuel temperature in the reactor are calculated (see reference numeral 11). The operation of the present embodiment is the same as that described in the above-mentioned [Operation], and the description thereof will be omitted.

According to this embodiment, in the node calculation performed using the nuclear constants obtained in the lattice calculation 4, the reactivity effect due to the difference between the output history assumed in the lattice calculation 4 and the output history of the actual machine is newly obtained. , The burning constants P _i and D _i for the Pu isotopes and Am isotopes indicated by reference numeral 6 from the nuclear constants Table 5 evaluated by the lattice calculation 4, the reactivity coefficients ∂K _∞ / ∂N _i ,
The microscopic cross-sectional area σ _i is used as tabular data.

Then, the change in atomic number density of Pu isotopes and Am isotopes due to the difference in the output history is evaluated in the core performance calculation, and the reactivity due to the difference in the output history is combined by combining the reactivity coefficient and the microscopic cross section. The effect can be corrected. Therefore, a highly accurate core performance estimation method can be obtained.

In the above embodiment, the combustion coefficient and the number of atoms of the Pu isotope and the Am isotope nuclide necessary for constructing the production rate and the decay rate in the combustion equation of the Pu isotope and the Am isotope nuclide in the small region are set. A means for creating the reactivity coefficient and the microscopic cross-sectional area necessary for obtaining the rate of change of the nuclear constant from the difference in density based on the tabulated data is added.

Further, by using the output history assumed when calculating the nuclear constant of the small region and the output history obtained by solving the node equation representing the neutron transport phenomenon of the entire core, Pu in the small region is calculated. Solving the combustion equations of isotopes and Am isotope nuclides, creating an atomic number density file 14 that registers the obtained two atomic number densities, and based on that, evaluate 9 the atomic number density difference ΔN _i due to the output history difference. .

Next, a second embodiment of the method for estimating core performance of a reactor according to the present invention will be described with reference to FIGS. 2 to 4. In FIG. 2, the same parts as those in FIG. 1 are designated by the same reference numerals and overlapping description will be omitted.

The difference between the second embodiment and the first embodiment is that the contents of the atomic number density file indicated by reference numeral 31 and the tabulated data indicated by reference numeral 32 in FIG. 2 are different. Then, when the amount of change in the nuclear constant in the small region is obtained, the amount of change in the nuclear constant caused by the difference in the decay amount of Pu-241 to Am-241 due to the β decay of the output history is obtained, The fifth means for correcting the described nuclear constant is added.

In this embodiment, the case of application to the first group node equation will be described. That, Pu-24 atomic density file 31 corresponding to the N ^a _Pu-241 (actual output history
Number 1 in atomic density), a N ^o _Pu-241 (atomic number density of Pu-241 that corresponds to the output history assumed in grid computing).

Further, the tabulated data 32 is And is registered as a function of the hysteresis moderator density (UH) obtained by averaging the burnup (E) and the moderator density by the burnup.

Of the 32 tabulated data, σ ^c
_Pu-240 and σ ^a _Pu-241 is, Pu-240 was used in combustion calculations performed in grid computing, Pu-241 multi-group microscopic cross section sigma ^{c for, g} _Pu-240 and σ ^{a, g} _Pu- An amount of ₂₄₁ energy-contracted into one group, Φ ₀ is the average neutron flux of the fuel assembly.

∂K _∞ / ∂N _Pu-241 and ∂K _∞ / ∂N
_Am-241 is a quantity that represents the amount of change in the infinite multiplication factor of the fuel assembly when the atomic number density of Pu-241 and Am-241 changes by a unit amount, and for each nuclide evaluated by the perturbation calculation in the lattice calculation. It is calculated by the following formula using the reactivity value of.

For the method of evaluating the reactivity value by the perturbation theory, for example, James J. Duderstat & Louis J. Hamilton “Nu
clear Reactor Analysis (1976) John Wiley & Sons, Inc.Ch
It is described in apter 7.

[0060]

[Equation 9]

Since the amount of Pu-241 produced during the operation of the reactor is large and the half-life of β-decay is relatively short at 14.4 years, the atomic number density changes greatly even if the burnup is the same when the output level changes. To do. On the other hand, Am-241 is Pu-24
It is a nuclide generated by β-decay of 1 and α-decays with a half-life of 432.7 years. Further, the two nuclides are both highly reactive.

In this embodiment, the infinite multiplication factor K ^m _∞ of the node m evaluated based on the kernel constant of the lattice calculation is used as the correction amount ΔK ^m _∞.
Add. The evaluation method of ΔK ^m _∞ will be described below. Pu-241 and Am-2 resulting from the difference in output history
If the atomic number density difference of 41 is ΔN _{Pu −241} and ΔN _{Am −241} ,

[Equation 10]

The method of calculating ΔN _Pu-241 and ΔN _Am-241 will be described below. The combustion equation of Pu-241 is expressed by the following equation.

[0064]

[Equation 11]

On the other hand, the combustion equation of Am-241 is expressed by the following equation.

[Equation 12] The symbols in equation (16) have the same meanings as the symbols appearing in equation (15).

When equations (15) and (16) are added,

[Equation 13]

Since Am-241 has a long half-life of 432.7 years, the final term λ _Am-241 N _{Am-2 41} of the equation (17) can be ignored.

Since the coefficient applied to the left side of the neutron flux Φ _{o in} equation (17) has a small power density dependence, Is determined by E and UH regardless of the reactor power level.

Therefore, since it can be approximated to ΔN _{Pu-241 to} ΔN _Am-241 , the equation (14) can be expressed as the following equation.

[0070]

[Equation 14]

The combustion equation (15) of Pu-241 is N
If the initial value of _Pu-241 is used as input data, (N ⁰ _Pu-241 σ ^c _Pu-240 Φ ₀ ) and (σ
You can solve it by using ^a _{Pu-2 41} Φ ₀ ).

Therefore, the Pu obtained by solving the equation (15) for the case of the actual output history and the constant output history assumed in the grid calculation is obtained.
-241 atomic number density N ^a _Pu-241 and N ⁰ _Pu-241 and tabulated data 32 (∂K _∞ / ∂N _Pu-241 -∂K _∞ / ∂
N _Am-241 ) can accurately evaluate the combustion change in the atomic number density of Pu-241 and Am-241 even when the output of the nuclear reactor changes, so that Pu-241 and Am-241 due to output fluctuations It is possible to correct the effect of the on the core reactivity.

Next, the effect of this embodiment will be described with reference to FIGS.
Will be described with reference to. Figure 3 shows a UO ₂ core loaded with an 8x8 array fuel with an initial average enrichment of 3.6 weight percent and an initial average Pu enrichment = 3.9 weight percent (U-235
Critical eigenvalue when the MOX core loaded with 8 × 8 array fuel (enrichment = 1.2 weight percent) is operated with the power density doubled at the rated power operation assumed in grid calculation. The conventional example (in the figure, referred to as the conventional method) is related to the evaluation error of
The present invention will be referred to as the present invention).

FIG. 4 shows the same UO ₂ core and M as in FIG.
In the OX core, the evaluation error of the critical eigenvalue in the case of 4-cycle operation during rated power operation, that is, when a half-year constant inspection period is set between each operation cycle, is compared with the conventional example (Fig. The present invention (in the drawings, referred to as the present invention) is shown in comparison.

As can be seen from FIGS. 3 and 4, UO ₂
In the case of the core, the evaluation error of the critical eigenvalue is small even when the operation greatly deviates from the rated output continuous operation,
In the case of MOX core, a large evaluation error occurs.

On the other hand, according to the embodiment of the present invention,
In the case of the MOX core, the evaluation error can be reduced to the same level as that of the UO ₂ core, and the improvement effect is recognized.

Next, a third embodiment of the method for estimating core performance of a nuclear reactor of the present invention will be described with reference to FIG. The difference of the present invention from the second embodiment is that the tabulated data 33 has N
⁰ _Eu-155 and ∂K _∞ / ∂NGd _-155 were changed, and the atomic number density file 31 was deleted. Here, an embodiment for improving the prediction accuracy of the cold temperature critical eigenvalue when starting each operation cycle will be described. When a nuclear reactor is operated, Eu-155 is produced in the fuel as a fission product, but Eu-155 is β-decayed with a half-life of 4.9 years to become Gd-155 which is a strong thermal neutron absorber.

The amount of Gd-155 produced as a fission product is Tasaka K. et al .: JNDC Nuclear Data.
Library of Fission Products- Second Version-, JAERI
-1320 (1990), the amount produced by β-decay of Eu-155 is overwhelmingly larger than the amount produced by Gd-155 directly.

In lattice calculation 4, since continuous combustion calculation is performed and each constant is evaluated, Eu generated as a fission product is calculated.
The effect that -155 is β-decayed during the regular inspection period and Gd-155 is generated cannot be considered by the conventional core performance estimation method.

The amount of Eu-155 transformed into Gd-155 by β decay during the regular inspection period is represented by the following equation.

[0081]

[Equation 15]

Eu-155 changed to Gd-155 during the regular inspection period.
The reactivity correction amount due to the decay into

[0083]

[Equation 16]

In this example, of the tabulated data 33, N ⁰ _Eu-155 was evaluated as N _{Eu -155 on} the right side of the equation (19), and ΔN _Gd-155 and tabulated data were evaluated.
By substituting ∂K _∞ / ∂N _Gd-155 in 33 into the equation (20), the correction amount of the infinite multiplication factor at each node can be obtained.

Second to fourth after the half-year regular inspection period is set between each operation cycle in the above-mentioned UO ₂ core and MOX core
The correction amount by the equation (20) for the cold core at the beginning of the cycle is −0.03% ΔK, −0 for the UO ₂ core, respectively.
In the case of 07% ΔK, -0.14% ΔK, and MOX core, each is -0.
It was estimated to be 04% ΔK, −0.08% ΔK, and −0.14% ΔK.
It can be seen that the effect of this example becomes remarkable in the high burnup core with many fission products.

Next, a fourth embodiment of the method for estimating core performance of a nuclear reactor of the present invention will be described with reference to FIG. Note that FIG.
In the figure, the same parts as those in FIG.

In this embodiment, the case of applying to the one-group node equation will be described. The difference between the fourth embodiment and the first embodiment is that the tabulated data 34 is not limited to the tabulated data 6 in the first embodiment, but the reactivity contributions to the Pu and Am isotopes are subtracted. The infinite multiplication factor K _∞ and the macroscopic cross-sectional area Σ are tabulated. These amounts are calculated by the following formula.

[0088]

[Equation 17] Further, the atomic number density file 35, only atomic number density N ^a _Pu-241 in radionuclide i corresponding to the actual output history is registered. Therefore, the atom number density ΔN _i due to the output history difference in FIG.
Evaluation 9 is unnecessary.

Correction 10 of each constant file used in the core performance calculation is performed by the following equation.

[Equation 18]

In this example, the atomic number density N ^a of the nuclide i corresponding to the actual output history is based on the infinite multiplication factor K _∞ and the macroscopic cross section Σ after subtracting the reactivity contributions to the Pu and Am isotopes. Since the correction is performed using _Pu-241 , the core constant in the case of burning the atomic number density of Pu isotopes and Am isotopes in the actual output history can always be used in the core performance calculation.

The effect of this embodiment is that the amount of data to be tabulated is slightly increased as compared to the data to be tabulated in the first embodiment, but the number of atoms of the nuclide i corresponding to the actual output history is recorded in the atom number density file. Since only the density needs to be stored, the memory of the computer can be greatly saved.

Next, with reference to FIG. 7, a first embodiment of the core performance estimating apparatus according to the present invention will be described. FIG. 7 is a block diagram showing an apparatus for carrying out the core performance estimation method for each reactor described above.

In FIG. 7, a neutron flux measurement system 41 and a state quantity measurement system 48 are installed above the core 47 in a core 47 in the nuclear reactor 46. The neutron flux measurement system 41 and the state quantity measurement system 48 Is connected to the input side of the data sampling device 42. The data sampling device 42 is connected to the process computer 43. An external storage device 44 and an input / output device 45 are connected to the process computer 43.

In this embodiment, the measurement systems 41 and 48 in the reactor 46 are
By inputting the count value from the above into the data sampling device 42 and using the core performance estimation method of each reactor described in the first to fourth embodiments.
The power density distribution in 47 can be calculated, and the thermal limit value of the core fuel can be displayed from this power density distribution.

That is, the reactor 46 is indispensable for evaluating the core performance such as the pressure inside the reactor 46, which is collectively represented as the state quantity measuring system 48 in FIG. 7, and the cooling flow rate of the core. A measuring instrument for measuring various quantities is installed, and the measured values are periodically transmitted to the process computer 43 by the data sampling device 42, or are transmitted at any time at the request of the process computer 43 side.

In the external storage device 44 of the process computer 43, data describing the shape of the core 47 in advance, nuclear constants representing the nuclear characteristics of the materials constituting the core 47, thermo-hydraulic constants of the core, and other physical property values are stored. Remember.

From the measured values of various quantities indispensable for core performance evaluation transmitted to the process computer 43 by the data sampling apparatus 42 and the quantities stored in advance in the external storage device 44, simultaneous linear homogeneous equation (1) Can be obtained. Further, the result evaluated by the process computer 43 may be transmitted to the output device 45.

Therefore, by applying the embodiment of the present invention to the inside of the process computer 43, a highly accurate on-line core performance estimation device for a nuclear reactor can be obtained.

Next, a second embodiment of the reactor core performance estimating apparatus of the present invention will be described with reference to FIG. In FIG. 8, reference numeral 51
Is a computer, and an external storage device 52 and an input / output device 53 are connected to the computer 51. In this embodiment, a future reactor operation plan is input to the computer 51 and the above-described reactor core performance estimation method is used.

With this, the power density distribution inside the core of the nuclear reactor is calculated, and the thermal limit value of the core fuel is calculated and displayed from the power density distribution, and at the same time, the eigenvalue of the simultaneous linear homogeneous equation (1) is calculated. It has a function of calculating an approximate value for and displaying it together with the critical adjustment parameter.

That is, the external storage device 52 of the computer 51 stores in advance data describing the shape of the core, the nuclear constants representing the nuclear characteristics of the materials forming the core, the thermohydraulic constants of the core, and other physical property values. I will let you.

Next, the operation plan of the future nuclear reactor is input from the input / output device 53 of the computer 51. Here, the future reactor operation plan is represented by numerical data such as a change in reactor output for a certain future period, a control rod insertion position, and a core flow rate. According to the present embodiment, a core performance estimation device with high accuracy can be obtained.

[0103]

According to the present invention, in the node calculation for evaluating the performance of a nuclear reactor using the nuclear constant obtained by the lattice calculation, the Pu isotope due to the difference between the output history assumed in the lattice calculation and the actual output history And the reactivity coefficient of the Am isotope without performing a new lattice calculation, the reactivity coefficient tabulated based on the lattice number calculation and the atomic number density of the Pu and Am isotopes calculated under the actual output history condition And microscopic cross-sectional areas can be combined to evaluate.

If this effect is taken into consideration in the prior art, a huge amount of calculation time is required because the grid calculation corresponding to the actual output history must be performed for each node. However, according to the present invention, a new grid calculation is required. The same effect can be taken into account in the core calculation without doing so.

Therefore, according to the present invention, it is possible to provide a highly accurate core performance estimation method and a core performance estimation apparatus using the same, which saves a great amount of calculation time.

[Brief description of drawings]

FIG. 1 is a first method of estimating core performance of a nuclear reactor according to the present invention.
6 is a flowchart showing an embodiment of FIG.

FIG. 2 is a second method of estimating core performance of a nuclear reactor according to the present invention.
6 is a flowchart showing an embodiment of FIG.

FIG. 3 compares the evaluation error of the critical eigenvalue when the power level of the reactor is operated at twice the power level assumed in the grid calculation between the conventional example and the present invention example, and the evaluation error when the present invention is applied. The characteristic view showing the improvement effect of.

FIG. 4 compares the evaluation error of the critical eigenvalue between the conventional example and the example of the present invention when the reactor is operated at the rated power level assumed in the grid calculation and a regular inspection period of half a year is provided between each operation cycle. FIG. 6 is a characteristic diagram showing the effect of improving the evaluation error when the present invention is applied.

FIG. 5 is a third method of estimating core performance of a nuclear reactor according to the present invention.
6 is a flowchart showing an embodiment of FIG.

FIG. 6 is a fourth method of estimating core performance of a nuclear reactor according to the present invention.
6 is a flowchart showing an embodiment of FIG.

FIG. 7 is a first core performance estimating apparatus for a nuclear reactor according to the present invention.
FIG.

FIG. 8 shows a second reactor core performance estimating apparatus according to the present invention.
FIG.

FIG. 9 is a flowchart showing a conventional core performance estimation method for a nuclear reactor.

[Explanation of symbols]

41 ... Neutron flux measurement system, 42 ... Data sampling device, 43
... Process computer, 44, 52 ... External storage device, 45, 53 ... Input / output device, 46 ... Reactor, 47 ... Core, 48 ... State quantity measurement system,
51 ... calculator.

Claims (8)

[Claims] 1. A reactor core is spatially divided into subregions, and boundary conditions are assumed at the outer boundaries of the subregions. Solving for combinations of parameters such as presence / absence, moderator density, fuel and moderator temperature, and solving the combustion equation of the nuclide assuming the output history of the small region, and calculating the nuclear constant of the small region in advance. Then, a nuclear constant table of infinite multiplication factor and macroscopic cross section is created, and by using this nuclear constant table, the node equation expressing the neutron transport phenomenon of the whole core of the reactor and the thermohydraulic behavior of fuel and moderator In the method of estimating the core performance of the reactor by solving the equation or empirical formula, the output history of the small area obtained by solving the node equation representing the transport phenomenon of neutrons of the entire core for each time step, The above When there is a difference between the output history assumed when solving the nuclide combustion equation in the small area, including the shutdown period, the Pu isotope and Am in the small area resulting from the difference with the actual output history. A first means for calculating the amount of change in the nuclear constants due to the difference in the atomic number density of the isotopes and correcting the nuclear constants described in the nuclear constant table, and the nuclear constant table described in the nuclear constant table Second means for taking in and correcting the nuclear constant by the first means, and using the corrected nuclear constant to calculate a coefficient of a node equation representing a neutron transport phenomenon of the entire core of a nuclear reactor, Third means for obtaining the power density distribution in the reactor by solving the node equation representing the neutron transport phenomenon by incorporating the corrected coefficient, and incorporating the power density distribution, at least the density of the moderator in the reactor and the Calculate fuel temperature Core performance estimation method of the reactor, characterized in that it consists of a fourth means. 2. The amount of change in the nuclear constant caused by the difference in the decay amount of Am-241 due to the β-decay of Pu-241 due to the difference in the output history is determined when the amount of change in the nuclear constant in the small region is obtained. The core performance estimating method for a nuclear reactor according to claim 1, further comprising: a fifth means for correcting the nuclear constants described in the nuclear constant table. 3. In determining the amount of change in the nuclear constant of the small region, the amount of change in the nuclear constant caused by the difference in the decay amount of Gd-155 due to β decay of Eu-155 due to the difference in the output history is obtained. The core performance estimation method for a nuclear reactor according to claim 1, further comprising: a sixth means for correcting the nuclear constants described in the nuclear constant table. 4. The amount of change in the nuclear constant is calculated from the difference between the fuel coefficient and the atomic number of the nuclide necessary to assemble the production rate and decay rate in the combustion equation for Pu and Am isotopes in the small region. The reactor core performance estimating method according to claim 1, further comprising a seventh means for creating a reactivity coefficient necessary for the above as a tabled data of a microscopic cross-sectional area. 5. The Pu isotope and Am from the nuclear constant table.
An infinite multiplication factor and a macroscopic cross section excluding the isotope's reactivity contribution were obtained, and these amounts were tabulated together with the data created by the seventh means, and the Pu isotope was calculated under the actual output history condition. And the atomic number density of the Am isotope, the reactivity coefficient and the microscopic cross section are taken to calculate the reactivity value and macroscopic cross section of the Pu and Am isotopes for each node. 2. The infinite multiplication factor and macroscopic cross-sectional area excluding the reactivity contributions of Pu isotope and Am isotope are used as the infinite multiplication factor and macroscopic cross-sectional area for each node. Method for estimating core performance of nuclear reactors. 6. The output history assumed when calculating the nuclear constants of the small region and the output history obtained by solving the node equation representing the neutron transport phenomenon of the entire core 2 was obtained by solving the combustion equation of the nuclide of
The method for estimating core performance of a nuclear reactor according to claim 1, further comprising a file in which one atomic number density is registered. 7. A state quantity measuring system and a neutron flux measuring system installed in a nuclear reactor, a data sampling device for inputting measured values of these measuring systems, a process computer connected to this data sampling device, and this process. An apparatus for estimating core performance of a nuclear reactor, comprising an external storage device and an input / output device connected to a computer. 8. A computer for inputting a nuclear reactor operation plan as input, data for describing the shape of the core beforehand connected to this computer, a nuclear constant representing the nuclear characteristics of the material constituting the core, and a thermal-hydraulic constant of the core. An apparatus for estimating core performance of a nuclear reactor, further comprising: an external storage device for describing required physical property values and the like; and an input / output device for inputting an operation plan of a future nuclear reactor connected to the computer.