JPH11258382A - Method for calculating reactor core performance of reactor - Google Patents

Abstract

PROBLEM TO BE SOLVED: To accurately achieve calculation while saving storage capacity and a calculation time by obtaining a differential coefficient regarding the degree of combustion of the atom density difference of Pu-241 based on the difference between assumed output and actual output history and integrating it regarding the degree of combustion. SOLUTION: After a boundary, condition 42 when calculating reactor core performance is inputted, an atom density difference 47 of Pu-241 between assumed output and actual output history is evaluated before the repeated calculation of a reactor core hear water power calculation 51 and reactor core neutron flux density/output density calculation 51. Therefore, first, N<0> Pu-241 , σ<a> Pu-241 , and ϕ0 that are the functions of degree of combustion E, deceleration material density U, and the like are calculated by a nucleus constant file 54. Then, it is judged whether node output 44 in an output history is zero or not. If it is not equal to zero, a differential coefficient 45 regarding the degree combustion of the atom density difference 47 of Pu-241 is calculated by an equation. Then, the deceleration material density U is calculated by a reactor core thermal water power calculation 48 and further a node nucleus constant 49 is calculated from the nucleus constant file.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method and an apparatus for calculating a core performance used for designing, monitoring, or managing the operation of a reactor core.
u) The present invention relates to a core performance calculation method and a core performance calculation device suitable for a core loaded with fuel.

[0002]

2. Description of the Related Art A reactor core performance calculation method or core performance for evaluating the behavior of neutrons in a core based on a physical model is used for designing a fuel / core of a nuclear reactor, planning an operation plan, and managing the operation. Computing equipment is required. In the core performance calculation, the core is spatially divided into small areas called nodes, and the nodes that can calculate the burnup of fuel, the temperature of fuel and moderator, the moderator density, the presence or absence of control rods, etc. as parameters for each node Given a nuclear constant, calculate the neutron behavior of the core based on neutron transport or diffusion theory.

[0003] The output (heat generation) distribution is calculated from the neutron distribution obtained by the calculation, and combined with the thermal hydraulic calculation for evaluating the heat removal performance of the fuel rod, the required performance can be obtained while maintaining the soundness of the core. Evaluate if it works. The required node nuclear constant is generally obtained by the following procedure.

First, two-dimensional neutron transport calculation (lattice calculation) is performed for each fuel assembly for each cross section where the fuel rod enrichment distribution, enrichment distribution, water rod arrangement, etc. in the fuel assembly differ. By solving the combustion equation of the nuclides contained in, the change in combustion of the atomic number density N _i and the microscopic cross section σ _i of each nuclide i is obtained. Here, in order to reduce information necessary for core performance calculation, the atomic number density N _i and the microscopic cross section σ _i
Instead, the macroscopic cross-section さ れ る defined by the following equation, which is the sum of the product of the microscopic cross-section of an isotope nuclide and the atomic number density for all nuclides, is often used as the node nuclear constant.

[0005]

(Equation 1)

[0006] In the grid calculation, a plurality of calculations are performed under a given aggregate output condition while changing the moderator density, temperature, fuel temperature, presence / absence of control rods, and the like. The node nuclear constant obtained in this way is fitted with the burnup and the above parameters, and a nuclear constant file storing fitting coefficients is created.
The fitting nuclear parameters include the macroscopic cross section Σ
In addition, the infinite multiplication factor and the diffusion coefficient are also included.

[0007] There are nuclides in the fuel whose atomic number density changes greatly with the time change of the reactor power, and the reactor reactivity based on the atomic number density difference caused by the difference between the power assumed in the lattice calculation and the actual power history. There are nuclides whose effects cannot be ignored in reactor design and operation management.

Xe-135 and S, which are strong neutron absorbers
m-149 is one such nuclide. For example, Xe
-135 is produced as a fission product and I-1
Also produced by β decay of 35.

On the other hand, Xe-135 absorbs neutrons to produce X
e-136, or Cs-135
Disappears by converting to. Since β decay is governed only by time, even if the burnup of fuel is the same, the atomic number density of Xe-135 changes according to the core power history, and the resulting change in reactivity cannot be ignored.

As a method of handling such nuclides, for example, in the Central Research Institute of Electric Power Industry's report 284006, “Analysis of core characteristics of 1.1 million-class BWR plant” (1984), it is assumed that atoms of equilibrium Xe-135 in assumed power state are used. Number density and actual X
A method is described in which a difference in the atomic number density of e-135 is calculated, and a macroscopic cross-sectional area used for core calculation is corrected by using the atomic number density difference of Xe-135 and a microscopic cross-sectional area. .
The actual atomic number density of Xe-135 is obtained by integrating a differential equation over time.

Other than Xe-135 and Sm-149, there are other nuclides having a great influence on the nuclear constant, and a method of specially treating them as a node nuclear constant has been proposed. For example, “Advanced Methodol” by CWLindenmeier et al.
ogy for LWR Design, "International Reactor Physics
At the Conference, Jackson Hole (1988), isotopes that greatly affect the node nuclear constant are completely separated and treated with a microscopic cross section, and the remaining nuclides are treated with a macroscopic cross section collectively. The method is described. The major isotopes of uranium, plutonium, and gadolinium are listed as the isotope nuclides to be handled separately, and the combustion change of the atomic number density is calculated for those nuclides.

Among the above plutonium isotopes,
Pu-241 converts from fissile material to Am-241 which is a neutron absorbing material by β decay with a half life of 14.4 years
Is particularly important from the viewpoint of reactivity change. “Critical Sizes of Light−Water Moderated UO ₂ and
PuO ₂ -UO ₂ Lattices “JAERI-1254 (1978) states that the decrease in Pu-241 due to β decay is equal to the increase in Am-241, and the reactivity due to the conversion of Pu-241 to Am-241. In recent years, uranium-plutonium mixed oxide fuel (MOX) for light water reactors has been evaluated.
Fuel) has been studied, and there has been a demand for a core performance calculation that can take into account the above-described reactivity change.

JP-A-6-86370, entitled "Method and Apparatus for Estimating Reactor Core Performance of Reactor" includes Pu-241 and Am
A specific example of a core performance calculation method for calculating the difference between the reference atomic number density and the actual atomic number density by focusing on -241 and correcting the infinite multiplication factor of the node based on the atomic number density difference is described. ing.

The present invention proposed here is the same as the technique disclosed in Japanese Patent Application Laid-Open No. 6-86370 in that the node nuclear constant is corrected based on the atomic number density difference and that attention is paid to Pu-241. , Pu-241 is characterized in that the difference between the reference atomic number density and the actual atomic number density can be obtained more accurately with less information than in the above-mentioned conventional technology. Here, in order to clarify the features of the present invention, a comparative Japanese Patent Application Laid-Open No.
The technique described in No. 6370 will be described.

[0015] the above-described conventional art, .DELTA.k the correction term of infinite multiplication factor is one of the nodes nuclear constants ^∞

[0016]

[Number 2] ^{Δk ∞ = (∂k ∞ / ∂N} Pu-241) ΔN Pu-241 Calculated by _{^{+ (∂k ∞ / ∂N Am-}} 241) ΔN Am-241 ... ( number 2). Here, ΔN _Pu-241 is the difference in atomic number density of Pu-241 caused by the difference between the assumed output history and the actual output history, and ΔN _Am-241 is the difference between the assumed output history and the actual output history. This is the difference in the atom number density of generated Am-241. Furthermore, as with Tsuruta et al., The relationship between ΔN _Pu-241 and ΔN _Am-241

[0017]

[Expression 3] ΔN _Pu-241 = −ΔN _Am-241 (Expression 3) is used.

ΔN _Pu-241 is the number density of Pu-241 corresponding to N _Pu-241 corresponding to the actual output history, and the number of atoms of _Pu-241 corresponding to the output history assuming ΔN ⁰ _Pu-241 in the lattice calculation. The density is calculated by the following equation.

[0019]

ΔN _Pu-241 = N _Pu-241 −ΔN ⁰ _Pu-241 (Equation 4) Atomic density N of Pu-241 corresponding to the actual output history
_Pu-241 is Pu-24 by neutron capture of Pu-240
The following equation, which takes into account the generation of 1 and the disappearance of Pu-241 due to neutron absorption and β decay, is obtained by time integration.

[0020]

(Equation 5)

Where N ⁰ _Pu-240 : Pu-240 atomic number density evaluated by lattice calculation σ ^c _Pu-240 : Pu-240 microscopic cross section evaluated by lattice calculation φ ₀ : By lattice calculation Neutron flux used σ ^a _Pu-241 : microscopic absorption cross section of Pu-241 evaluated by lattice calculation λ _Pu-241 : decay constant of Pu-241 (= 1.5310 ×
10 ^-9 sec = 1.3228 × 10 ^-4 days) P: actual node output P _0: is assumed node output grid computing.

[0022] _{^{(∂k ∞ / ∂N Pu-241}} ), (∂k ∞ / ∂N
_Am-241 ), N ⁰ _Pu-241 , N ⁰ _Pu-240 , σ ^a _Pu-241 , σ ^c
_Pu-240 and φ ₀ are stored in the nuclear constant file as functions such as burn-up and moderator density, and λ _Pu-241 and P ₀ are also provided as input data to the core performance calculator.

[0023]

Since the core performance calculation device requires a high calculation speed, it is preferable that the number of reference nuclear constant files be small. In the above technique, Pu
-241 to calculate the atomic number density difference ΔN _Pu-241
In addition to the atomic number density N ⁰ _Pu-241 of Pu-241 and the microscopic absorption cross section σ ^a _{Pu-241 of Pu-241} corresponding to the output history assumed in the lattice calculation, the microscopic capture of Pu-240 A nuclear constant file concerning the area σ ^c _Pu-240 and the atomic number density N ⁰ _Pu-240 of _Pu-240 evaluated by lattice calculation is required.

Further, when calculating the atomic number density difference ΔN _Pu-241 of _Pu-241 using (Equation 5), fuel is loaded into the core in order to input the output of the node with respect to time, that is, the output history. The Pu-241 atomic number density difference ΔN _Pu-241 can be evaluated when the power history after that is different from the planned power history, but ΔN when the combustion start time of the new fuel newly loaded in the core is earlier than expected. _Pu-241 could not be calculated.

[0025]

According to the present invention, the differential coefficient d (ΔN) relating to the burnup E of the atomic number density difference ΔN _Pu-241 of _Pu-241 based on the difference between the assumed output and the actual output history.
_Pu-241 ) / dE is calculated and ΔN _Pu-241 is calculated by integrating it with respect to the burnup. By this method, Pu-240
The nuclear constant file concerning the microscopic absorption cross section σ ^c _Pu-240 and the atomic number density ΔN ⁰ _Pu-240 evaluated by the lattice calculation becomes unnecessary.

For the fuel initially loaded in the reactor core, a time difference between the scheduled start time of fuel combustion and the actual start time of fuel combustion or time information capable of calculating the time difference is input, and one cycle has already been performed. For fuels that have been loaded into the core above, the zero output period can be obtained by inputting the date and time difference between the combustion end date and time of the previous cycle and the combustion start date and time of the current cycle or the time information for calculating the date and time difference. , The atomic number density difference ΔN _Pu-241 of _Pu-241 can be easily calculated.

[0027]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS A first embodiment of a method for calculating core performance of a nuclear reactor according to the present invention will be described below with reference to FIG. FIG. 1 shows Pu-24 as the nuclide of interest in the present invention.
1 is a block diagram showing an outline of a core performance calculation method for a MOX core in which No. 1 is selected, and a portion surrounded by a dotted line is disclosed in
This is a feature technology of the present invention that is significantly different from the conventional technology described in Japanese Patent Application No. 86370. In this example,
The number of neutron energy groups in the core performance calculation is one,
Alternatively, a modified first group model is used.

In the core performance calculation, first, initial state data 41 such as core shape data, core burnup distribution, core power distribution, and moderator density distribution in the core are input. If there is no accurate initial state data on the core power distribution or the moderator density distribution in the core, the estimated value may be used. Next, the boundary conditions 42 in the core performance calculation such as the core flow rate, the reactor pressure, the core inlet coolant temperature, and the control rod insertion depth are input.

In the core performance calculation for the uranium core, an iterative calculation of the core thermal hydraulic calculation 48 and the core neutron flux distribution and power distribution calculation 51 is subsequently performed. And the actual power history, the atomic number density of Pu-241 causes a large difference in reactivity between the planned and actual uranium cores. Then, the atomic number density difference 47 (= Δ
N _Pu-241 ). For this purpose, in the present invention, first, N ⁰ _Pu-241 , σ ^a _Pu-241 , which are functions of the burnup E, the moderator density U, the history moderator density UH which is the moderator density averaged by the burnup, and the like. φ ₀ is calculated using the nuclear constant file 54 created from the lattice calculation results. JP-A-6-863
In the technology of No. 70, N ⁰ _Pu-240 and σ ^c _Pu-240 were also required, but are not required in the present invention. Note that the above grid calculation result is
After the fuel has been prepared, it is obtained under the calculation conditions of loading the reactor core after a certain period of time and burning at a constant power density as a reference.

Next, it is determined whether or not the node output 44 (= P) in the actual output history is zero.

When the node output 44 is not zero, a differential coefficient 45 (= dΔN _Pu-241 / dE) relating to the burnup of the atomic density difference 47 of Pu-241 between the plan and the actual is calculated. The differential coefficient 45 can be calculated by the following equation.

[0032]

(Equation 6)

Here, (dE / dt) ₀ is the burnup per unit time in the grid calculation, and the other variables are defined in the same way as in (Equation 4) and (Equation 5). (DE / d
t) ₀ and the node output P ₀ in the lattice calculation can be calculated by the following equation.

[0034]

( _DE / dt) ₀ = P ₀ / M _hv

[0035]

P ₀ = PD ₀ × V where M _hv is the total heavy metal mass of the node, PD ₀ is the power density assumed in the lattice calculation, and V is the node volume.

(Equation 6) is, for example, to calculate the burn-up increment by Δ
E can be solved by the Euler integration method shown by the following equation.

[0037]

(Equation 9)

At this time, in order to improve the evaluation accuracy, the σ ^a _Pu-241 and φ ₀ and N ⁰ _Pu-241 are fitted using the nuclear constant file 54 during integration to the target burnup. You can start again.

Equation (6) cannot be operated when the node output is zero, because the denominator has the node output P. Therefore, when the node output is zero or small enough to be regarded as sufficiently zero, the following equation is obtained by analytically integrating the equation where the node output P is zero in (Equation 5).

[0040]

ΔN _Pu-241 = ΔN ^* _Pu-241 − (N ⁰ _Pu-241 + ΔN ^* _Pu-241 ) exp (−λ _Pu-241 × T) (Expression 10) where ΔN ^* _Pu-241 is , The difference between the atomic number density of Pu-241 between the plan before correcting the atomic number density during the period when the output is zero and the actual number density. T is a period of zero output in which a difference between the plan and the actual occurs.

That is, for the fuel to be loaded into the reactor for the first time, it is the time difference between the actual combustion start date and time and the scheduled combustion start date and time. On the other hand, it is the date and time difference between the combustion start date and time of the current cycle and the combustion end date and time of the previous cycle. Therefore, T has a positive value for fuel that has been loaded into the core for one cycle or more, but T can take any of positive and negative values for fuel initially loaded.

After obtaining ΔN _Pu-241 by the above operation,
The moderator density U is calculated by the core thermal hydraulic calculation 48. Further, a node nuclear constant 49 such as an infinite multiplication factor, a diffusion coefficient and a nuclear cross-sectional area is calculated from the nuclear constant file using the burnup, the moderator density and the like as parameters. Correction nodes nuclear constants 49 based on .DELTA.N _Pu-241 is in this embodiment performed on the infinite multiplication factor k ^∞. More specifically, in advance to prepare the nuclear constant file in advance under the approximation of the (number _{^{3) (∂k ∞ / ∂N Pu}} -241 -∂k ∞ / ∂N Am-241), ( number 2 ) by calculating a correction term .DELTA.k ^∞ infinite multiplication factor. The core neutron flux distribution and the power distribution calculation 51 are calculated using the corrected node nuclear constant, and the calculation is repeated until the thermal hydraulic calculation and the nuclear calculation converge. After convergence, the thermal margin is calculated, and the new burnup and hysteresis moderator density 53 are calculated. When the burnup has not reached the burnup at which the calculation is completed, the process returns to the setting part of the new boundary condition 42 and the calculation is continued.

Equation (6) can be derived by the following procedure. Now, as shown in FIG. 4, a time when a constant node output in the lattice calculation (reference output) P ₀ is burned from the burnup E until burnup E + &Dgr; E respectively _{t 0, t 0 + δt 0} , N
_The amount of change of _Pu-241 is δN ⁰ _Pu-241 . On the other hand, when the combustion is performed from the burnup E to the burnup E + δE at the actual output P (t), the change amounts of t, t + δt, and N _Pu-241 are δN _Pu-241 , respectively. Other shall assign a zero index in the physical amount when burned at the reference output P _0. The following equation holds from the relationship between burnup, power density, and time.

[0044]

(DE / dt) = P / P ₀ × (dE / dt) ₀ (Equation 11) Since the atomic number density of Pu-240 depends on the burnup, it hardly depends on the output history. The atomic number density change dN ⁰ _Pu-241 / dE of Pu-241 with respect to the burnup when burning at the reference output P ₀ can be expressed by the following equation using (Equation 4).

[0045]

## EQU12 ## dN ⁰ _Pu-241 / dE = ｛N ⁰ _Pu-240 σ ^c _Pu-240 φ ₀ −N _Pu-241 (σ ^a _Pu-241 φ ₀ + λ _Pu-241 )｝ / (dE / dt) ₀ (Equation 12) On the other hand, the neutron flux φ in the actual output history and the neutron flux φ in the reference output history ₀ for the same burnup E

[0046]

(Equation 13)

It is approximated that the above relationship is established, and the following equation is obtained as the change in the atomic number density of _Pu-241 dN _Pu-241 / dE with respect to the burnup when burning at the actual output.

[0048]

[Equation 14]

Here, taking the difference between (Equation 14) and (Equation 12),

[0050]

DN _Pu-241 / dE-dN ⁰ _Pu-241 / dE = Δ (dN _Pu-241 / dE) = d (ΔN _Pu-241 ) / dE (Equation 15) and (Equation 11) Is used, the term of Pu-240 is eliminated, and (Equation 6) can be derived.

In the above embodiment, the neutron energy structure in the nuclear calculation is one group or one corrected group. However, this method can be applied to multi-group calculation of two or more groups.

In another embodiment using the multi-group nuclear constant, a differential coefficient 45 (= dΔN _Pu-241 / dE) of the atomic number density difference 47 of Pu-241 between the plan and the actual is calculated by (Equation 16). .

[0053]

(Equation 16)

The difference between (Equation 6) and (Equation 16) is that the microscopic absorption cross section σ ^{a, g} _Pu-241 and the neutron flux φ ^g ₀ have an energy group structure.

For this reason, such information is stored in the nuclear constant file 54 in advance. When the multi-group nuclear constant is used, the structure of the neutron flux energy distribution becomes detailed, so that the accuracy is improved as compared with the one-group nuclear constant.

[0056] by integrating with the (number 16) for example (9) calculates the ΔN _Pu-241, multigroup nuclear constant for the reactor core performance calculation by substituting .DELTA.N _Pu-241 in (Equation 17) Can be corrected.

[0057]

[Equation 17]

Here, Σ ^{j, g} : macroscopic cross-sectional area of energy group g for nuclear reaction j (nuclear reaction j is, for example, neutron absorption reaction or formation reaction) (Σ ^{j, g} ) ^{0 *} : calculated by lattice calculation Pu-241 and Am-
Macroscopic cross-sectional area of energy group g for nuclear reaction j excluding the contribution of 241 N ⁰ _Am-241 : Am-241 atomic number density evaluated by lattice calculation σ ^{j, g} _Pu-241 : Evaluated by lattice calculation Σ ^{j, g} _Am-241 : microscopic cross section of energy group g for nuclear reaction j of Am-241 evaluated by lattice calculation (Σ ^{j , g} ) ^{0 *} , σ ^{j, g} _Pu-241 , σ ^{j, g} _Am-241 are stored in advance in the nuclear constant file as a function of burnup E, moderator density U, history moderator density UH, etc. deep.

As another embodiment for correcting the multi-group nuclear constant used in the core performance calculation, as shown in (Equation 18), the macroscopic cross-sectional area (格子^{j, g} ) ⁰ calculated by the grid calculation is calculated by the grid calculation. And the actual atomic number density difference between _Pu-241 and ΔN _Pu-241 and Pu-
The correction can also be made using the microscopic cross-sectional areas of the H.241 and Am-241.

[0060]

(Equation 18)

Also in this case, (Σ ^{j, g} ) ⁰ , σ ^{j, g} _Pu-241 , σ
^{j, g} _Am-241 is previously stored in the nuclear constant file as a function of the burnup E, the moderator density U, the history moderator density UH, and the like.

The method using the multi-group nuclear constant has the advantage of improving the accuracy, but the number of nuclear constants to be referenced increases, so that the calculation speed becomes slow. Therefore, ΔN is calculated by using only the nuclear constant of the heat group, which has the largest influence, among the multi-group nuclear constants in (Equation 16).
It is also possible to calculate _Pu-241 and correct only the heat group in the macroscopic cross section. In addition, the correction of (Equation 17) and (Equation 18) can be performed only for the nuclear constant of the heat group that has a large influence. By doing so, the evaluation accuracy can be improved over the one-group calculation while preventing the calculation speed from lowering.

The effect of the present invention will be described with reference to FIG. FIG.
Is an M loaded with 3.7% initial average enrichment MOX fuel
For the OX core, the eigenvalue evaluation errors of the case where the atomic number density of Pu-241 was not corrected at all and the case where the atomic number density of Pu-241 was corrected by the multigroup nuclear constant according to the present invention were shown. Things. Here, the MOX fuel of the first cycle is loaded into the core one year earlier than planned,
After the cycle, there was a six-year outage between cycles, and the newly loaded fuel was charged as planned. The detailed calculation was evaluated using the nuclear constant evaluated by the lattice calculation based on the above history.

Focusing on the case where the atomic number density of Pu-241 is not corrected at all, in the first cycle, since the fuel loading is one year earlier than planned, if the same control rod and flow pattern as the detailed calculation conditions are given, The reactivity was underestimated. The evaluation error changed as the combustion progressed. Conversely, the reactivity is overestimated in the second cycle as compared with the detailed calculation, and the difference is widened in the third cycle.

On the other hand, when the present invention is applied, the evaluation error is reduced.
1 is 1/4 or less of the case where the correction of the atomic number density is not performed. In this example, the newly loaded fuel is loaded as planned after the second cycle, but the present invention can evaluate the effect even if the loaded fuel after the second cycle has a difference from the plan. . This calculation does not require information on the microscopic absorption cross section σ ^c _{Pu-240 of Pu-240} or the atomic number density N ⁰ _Pu-240 evaluated by lattice calculation. For this reason, the storage capacity of data can be saved and the calculation time can be reduced.

In the present invention, instead of inputting the output history, for the fuel initially loaded in the core, the difference between the scheduled start date and time of the fuel combustion and the actual start date and time of the fuel combustion,
For fuels that have been loaded into the core for one or more cycles, the difference between the combustion end date and time of the previous cycle and the combustion start date and time of the current cycle is captured, and Pu-24 is calculated by (Equation 10).
The difference in atomic number density of 1 is evaluated. As a result, even if the fuel loading in the first cycle is earlier than planned, it is possible to evaluate the difference in atomic number density from the time when Pu-241 was planned.

Next, with reference to FIG. 2, one embodiment of a reactor core performance calculating apparatus according to the present invention will be described. FIG. 2 is a block diagram of a core performance calculation device that realizes the core performance calculation method described above.

The neutron flux measuring system 1 is installed in the core 2 in the reactor 1.
3 are installed. A control rod 3 can be inserted into the core 2,
A control rod position measuring system 14 is provided. In addition, temperature measurement system 12 for measuring reactor water temperature, pressure measurement system 1
0, flow measurement system 1 for measuring recirculation flow rate, main steam flow rate, etc.
1, and these measurement systems are connected to the data sampling device 20. The data sampling device 20 is connected to a process computer 21 having a built-in core performance calculation device 24. The process computer 21 has an input / output device 23
And the external storage device 22 are connected. A part of the output of the process computer 21 is input to the control device 25.

Measurement values such as pressure, flow rate, temperature, neutron flux and control rod position obtained from the measurement systems 10 to 14 in the reactor 1 are collected by the data sampling device 20 and transmitted to the process computer 21 periodically. You. The data is transmitted from the data sampling device 20 in response to a request from the process computer 21.

The process computer 21 has an input / output device 23
Before starting the reactor, the nuclear constant file 54 describing the shape data of the core 2 and the nuclear characteristics of the fuel constituting the core 2, data representing the thermal-hydraulic characteristics of the core, physical data of water-steam, etc. And the data is stored in the process computer 21
Is stored in the external storage device 22. Further, an atomic number density difference information input file necessary for utilizing the technology of the present invention is also input from the input / output device 23 and stored in the external storage device 22. The embodiment of the present invention shown in FIG. 1 is applied inside the core performance calculation device 24, and the power distribution and thermal margin of the core are calculated online, and the input / output device 23 and the control device 2
5 is output.

FIG. 6 is a schematic diagram showing an example of the structure of the atomic number density difference information input file unique to the present invention. In this example, the assembly numbers 1, 4, and 7 are the initially loaded fuels, the assembly numbers 2 and 5 are the fuel in the second cycle, and the assembly numbers 3 and 6 are 3
It is the cycle fuel. The fuel type identification number and the time information 1 are stored for each assembly, and the atomic number density difference from the lattice calculation of Pu-241 is stored for each fuel axial node (here, 24 nodes).

Here, the time information 1 is the planned combustion start time for the initially loaded fuel, and in this example, it is midnight on May 5, 1997. The time information 1 of the fuel in the second and third cycles is the end time of the combustion in the previous cycle. In this example, it is 22:00 on February 23, 1996. As for the atomic number density difference of Pu-241, 0 is input to all nodes for the initially loaded fuel, and the difference from the lattice calculation time is stored for each node in the second and third cycle fuels. . Further, the combustion start time of the reactor is input as the time information 2. This information is given as input data in advance, or is input from a timer of the process computer 21. In this example, the time information 2 is 6:00 on May 21, 1996.

The embodiment of FIG. 1 is applied to the inside of the core performance calculation device 24 built in the process computer 21 and the information stored in the above-mentioned atomic number density difference information input file is used as (Expression 10). The MOX core is calculated by calculating the atomic number density difference of Pu-241 based on the zero output period, and calculating the atomic number density difference of Pu-241 by (Equation 6) and (Equation 9) during burning. , A highly accurate core performance calculation device can be obtained. Further, the present invention is not limited to the MOX core,
Although the effect of improving accuracy is small, it can be applied to conventional uranium cores.

The atomic number density difference information input file is stored in the external storage device 22. The external number storage device 22 stores the atomic number density difference information output file for inheriting the atomic number density difference information necessary for the calculation of the next cycle. Is newly created. FIG. 7 is a schematic diagram showing an example of the structure of an atomic number density difference information output file. When the reactor heat output is high enough,
An atomic number density difference information output file is created. FIG.
This is an example of the contents of an atomic number density difference information output file when the reactor power reaches 100 MW on May 21, 996. Both the time information 1 and the time information 2 store the current time, and the atomic number density difference information calculated based on the present invention is stored for each node. This file is updated as time advances. When the reactor is shut down, an atomic number density difference information input file for the next cycle can be created based on the atomic number density difference information output file.

In the above embodiment, for the fuel initially loaded in the core of the nuclear reactor, the scheduled start date and time of the fuel combustion and the actual start date and time of the fuel combustion have already been loaded into the core for one or more cycles. For fuel, the time difference between the combustion end date and time of the previous cycle and the combustion start date and time of the current cycle is input as time information 1 and time information 2, but the time difference between time information 1 and time information 2 is input. But it is good. At this time, the information to be input is reduced by one.

FIG. 3 is a block diagram showing another embodiment of the core performance calculating apparatus according to the present invention. This device calculates the core performance off-line for the design and operation planning of the reactor core. An input / output device 33 and an external storage device 32 are connected to the digital computer 31. The embodiment of FIG. 1 of the present invention is applied to the inside of the digital computer 31, a nuclear constant file describing the shape data of the core 2 and the nuclear characteristics of the fuel constituting the core 2 from the input / output device 33,
Data representing the thermal-hydraulic characteristics of the reactor core, water-steam physical property data, and the like are input and stored in the external storage device 32. Further, an atomic number density difference information input file necessary for utilizing the technology of the present invention is also input from the input / output device 33 and stored in the external storage device 32. With such a configuration,
It is possible to obtain a high-precision core performance calculation device that realizes the design of the MOX core and the drafting of an operation plan offline.

[0077]

As described above, according to the present invention, M
Changes in reactivity of the OX core can be evaluated without using a nuclear constant file related to the atomic number density of Pu-240 evaluated by the microscopic absorption cross section of Pu-240 and lattice calculation, saving storage capacity and calculation time Thus, a highly accurate core performance calculation method and a core performance calculation device can be provided.

Further, by changing the calculation method of the atomic number density difference during the period of zero output from when the output is output, even if the combustion timing of the initially loaded fuel is earlier than the scheduled combustion start date, the influence is also reduced. The core performance calculation taking into account can be realized.

[Brief description of the drawings]

FIG. 1 is a block diagram showing an embodiment of a core performance calculation method for a MOX core in which Pu-241 is selected as a target nuclide in the present invention.

FIG. 2 is a block diagram showing an embodiment of a core performance calculation device according to the present invention.

FIG. 3 is a block diagram showing another embodiment of the core performance calculation device according to the present invention.

FIG. 4 is a schematic diagram showing the relationship between time and burnup during grid calculation and actual output history used in the present invention.

FIG. 5 shows the effect of improving the eigenvalue evaluation error of the reactor between the case where the atomic number density of Pu-241 is not corrected at all and the case where the atomic number density of Pu-241 is corrected by the multigroup nuclear constant according to the present invention. FIG.

FIG. 6 is a diagram of an embodiment showing file contents of an atomic number density difference input file used in the present invention.

FIG. 7 is a diagram of an embodiment showing file contents of an atomic number density difference output file output in the present invention.

[Explanation of symbols]

REFERENCE SIGNS LIST 1 reactor, 2 core, 3 control rod, 10 pressure measurement system, 11 flow measurement system, 12 temperature measurement system, 13 neutron flux measurement system, 14 control rod position measurement system, 20 data Sampling device, 21 ... Process computer, 22, 32 ...
External storage devices, 23, 33 input / output devices, 24 core performance calculation devices, 25 control devices, 41 core shape data,
Initial state data, 42: Boundary condition data, 43: Calculation of data necessary for calculation of atomic number density difference ΔN _Pu-241 , 44: Node output, 45: Burnup derivative of atomic number density difference, 46:
Atomic number density difference ΔN based on burnup derivative of atomic number density difference
_Pu-241 , 47 ... Difference in atomic number density ΔN based on zero output period
_Pu-241 , 48 ... core thermal hydraulic calculation, 49 ... node nuclear constant,
50: Nuclear constant correction based on ΔN _Pu-241 , 51: Calculation of neutron flux distribution and output distribution, 52: Calculation of thermal margin, 53: Calculation of new burnup, history moderator density, 54: Nuclear constant file.

Continued on the front page (72) Inventor Kazuya Ishii 7-2-1, Omika-cho, Hitachi City, Ibaraki Pref. Hitachi, Ltd.

Claims (6)

[Claims] 1. The core of a nuclear reactor is spatially divided into small areas called nodes, and the burnup of fuel, the temperature of fuel and moderator, the density of moderator, the presence or absence of control rods, and the like are used as parameters for each node. It has a means to evaluate the neutron behavior of the core by giving a computable node nuclear constant, and for specific isotope nuclides,
In a reactor core performance calculation method having a means for correcting the node nuclear constant based on a difference between an atomic number density when burning at a power assumed in advance and an atomic number density when burning with an actual output history, The above atomic number density difference,
A method for calculating core performance of a nuclear reactor, wherein a derivative of an atomic number density difference relating to the burnup is integrated and calculated with respect to the burnup. 2. The reactor core is spatially divided into small areas called nodes, and the burnup of fuel, the temperature of fuel and moderator, the moderator density, the presence or absence of control rods, and the like are used as parameters for each node. It has a means to evaluate the neutron behavior of the core by giving a computable node nuclear constant, and for specific isotope nuclides,
In a reactor core performance calculation method having a means for correcting the node nuclear constant based on a difference between an atomic number density when burning at a power assumed in advance and an atomic number density when burning with an actual output history, The above atomic number density difference,
When the node output can be regarded as zero or zero, the time derivative of the atomic number density difference is calculated by time integration, and at other times, the derivative of the atomic number density difference of the nuclide is integrated with respect to the burnup. A method for calculating a core performance of a nuclear reactor, wherein the core performance is calculated. 3. The reactor core is spatially divided into small areas called nodes, and the burnup of fuel, the temperature of fuel and moderator, the moderator density, the presence or absence of control rods, and the like are used as parameters for each node. It has a means to evaluate the neutron behavior of the core by giving a computable node nuclear constant, and for specific isotope nuclides,
In a reactor core performance calculation method having a means for correcting the node nuclear constant based on a difference between an atomic number density when burning at a power assumed in advance and an atomic number density when burning with an actual output history, For the fuel loaded in the core for the first time, the difference between the assumed atomic number density and the actual atomic number density of the above nuclide is evaluated based on the time difference between the scheduled start time of fuel combustion and the actual start time of fuel combustion. For fuels that have already been loaded into the core for more than one cycle, the difference in atomic number density is evaluated based on the time difference between the combustion end date and time of the previous cycle and the combustion start date and time of this cycle. A method for calculating core performance of a nuclear reactor, the method comprising: 4. At least PU-241 is included as a specific isotope nuclide obtained by integrating the atomic number density difference between the case of burning with a power assumed in advance and the case of burning with an actual output history with respect to the burnup. The method for calculating core performance of a nuclear reactor according to any one of claims 1 to 3, wherein: 5. For a fuel initially loaded in the core of a nuclear reactor, a date and time difference between a scheduled start date and time of fuel combustion and an actual start date and time of fuel combustion or time information capable of calculating the date and time difference is input. For fuel that has already been loaded into the core for one or more cycles, the time difference between the combustion end date and time of the previous cycle and the combustion start date and time of the current cycle, or time information for calculating the date and time difference, is input. Reactor core performance calculation method. 6. The reactor core is spatially divided into small regions called nodes, and the burnup of fuel, the temperature of fuel and moderator, the moderator density, the presence or absence of control rods, and the like are used as parameters for each node. It has a means to evaluate the neutron behavior of the core by giving a computable node nuclear constant, and for specific isotope nuclides,
In a reactor core performance calculation method having a means for correcting the node nuclear constant based on a difference between an atomic number density when burning at a power assumed in advance and an atomic number density when burning with an actual output history, A method of calculating core performance of a nuclear reactor, wherein a calculation method of the atomic number density difference of the nuclide is changed between when the node output is zero or can be regarded as zero and at other times.