JPH1010275A - Core performance calculator - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a core performance calculator capable of accurately calculating the nuclear characteristics in core considering instantaneous and historical effect for the case of neighboring off-set bundles in D-grid core or during the distortion of channel boxes. SOLUTION: An assembly average nuclear constant calculation means 11 calculates assembly average nuclear constant evaluate in advance, by a multiple assembly detailed calculation in accordance with combination of neighboring D-grid bundle and off-set bundle or assembly average nuclear constant using a correction quantity for infinite grid average constant. An in-core power distribution calculation means 12 calculates a critical proper value, neutron flux distribution and power distribution in the core based on a core diffusion calculation using the nuclear constant calculated with the assembly average nuclear constant calculation means.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a core performance calculation device for a boiling water reactor.

[0002]

2. Description of the Related Art The core of a boiling water reactor is constructed by inserting a plurality of fuel assemblies (bundles). FIG. 15 shows a partially cutaway perspective view of the fuel assembly 1. FIG.
As shown in FIG. 1, the fuel assembly 1 is configured by bundling a plurality of fuel rods 3 in a channel box 2. In the core, the cooling water enters from the lower part of the core, so that the cooling water flows from the lower part to the upper part in the channel box 2. The cooling water flows between the fuel rods 3 in the channel box 2, receives heat from the fuel during that time, boils, and exits from the upper part of the core.

FIG. 16 is an explanatory view of the arrangement of the fuel assemblies 1 in the core. As shown in FIG. 16, between the channel boxes 2a to 2d of the four adjacent fuel assemblies 1, there is a water gap 4 through which non-boiling cooling water flows. Channel boxes 2a to 4 of fuel assemblies 1
A control rod 5 is provided as necessary at the position of the water gap 4 at the corner portion sandwiched between 2d, and at the diagonal position of the water gap 4 at the corner portion, a control rod for the in-core neutron flux measuring device is provided. An instrumentation tube 6 is provided.

Here, the initial boiling water reactor is shown in FIG.
As shown in FIG. 1, a so-called D-lattice fuel core, in which the water gap width LW on the control rod 5 side is wider than the water gap width LN on the in-core neutron flux instrumentation tube side, is employed. In such a D-lattice fuel assembly, the neutron moderating effect in the wide water gap LW (hereinafter, wide gap) is large, so that the thermal neutron flux on the wide gap LW side is smaller than that on the narrow water gap LN (hereinafter, narrow gap) side. As a result, the thermal neutron flux distribution in the fuel assembly 1 becomes asymmetric. Therefore, in order to flatten the fuel rod local power distribution, the wide gap LW
It is necessary to lower the enrichment of the fuel rod 3 on the side and to increase the enrichment on the side of the narrow gap LN.

Thus, if the enrichment distribution is not symmetric,
For example, when the channel box sways during an earthquake, the thermal neutron flux distribution may shift to the high enrichment side due to a change in the water gap width. In such a case, reactivity may be applied and lead to a scrum. To avoid this,
It is conceivable to make the width of the wide gap LW and the narrow gap LN close to each other in the D lattice fuel core. This is realized by decentering the center of the channel box 2 of the D lattice bundle toward the control rod 5 within a range that does not interfere with the control rod 5. The D lattice bundle in which the center of the channel box 2 is decentered toward the control rod 5 is called an offset bundle.

The offset is realized by replacing the D lattice bundle with the offset bundle step by step in each cycle. FIG. 18 shows a stage where the two D lattice bundles surrounding the control rod 5 are offset bundled. In FIG. 18, the channel box 2 of the fuel assembly 1 is shown.
b and 2c show the offset.

In the core performance calculation in a nuclear reactor, a neutron flux distribution and a power distribution in the core are calculated based on a three-dimensional nuclear thermohydraulic coupled physical model of the core. The fuel rod power normalized in the cross section of the assembly is called local power, and the thermal limit value of the core is calculated from the average output of the assembly cross section and the local power of the fuel rod. The maximum linear power density of the assembly is defined as the product of the maximum fuel rod local output peaking coefficient of the assembly cross section and the average output of the assembly cross section. The critical power ratio, which is a monitoring index of burnout of the fuel rods, is calculated as a function of the fuel rod R factor representing the pattern of the local power distribution around each fuel rod, the average power of the assembly, and the channel flow rate. .

For a method of calculating these thermal limit values, see, for example, the document “Three-dimensional BWR core simul
ator, "JAWooly, Licensing topica1 report, NEDO-20
953, 1976, General E1ectric Company.

Further, in order to improve the accuracy of a three-dimensional nuclear thermo-hydraulic coupled physical model of a reactor core, for example, the reference “TARMS: An
on‐Line Boiling Water Reactor Management System B
asedon Core Physics Simulator, "M. Tsuiki et al., Pro
ceedings of a topical meeting on Advances in React
or Computations, Salt Lake City, 1983, it is possible to improve the accuracy by correcting the neutron flux distribution in the core by core calculation using the measured and calculated values of the neutron flux detector in the core. Is being done.

Further, in the core performance calculation, the irradiation amount of the neutron detector is integrated in order to monitor the life of the neutron detector. Similarly, the control rod irradiation amount is integrated in order to monitor the life of the control rod.

In the three-dimensional calculation of the core, the critical eigenvalue of the core, the neutron flux distribution in the reactor, and the neutron flux in the reactor are calculated by the coarse mesh node method based on the neutron diffusion theory, using the average nuclear constant of the nodes obtained by homogenizing the respective assemblies. Calculate the internal power distribution.
The average nuclear constant of the aggregate is generally calculated by a detailed calculation in an infinite lattice system to which a reflection boundary condition is applied at the boundary of each aggregate. In a normal core, the water gap width of the adjacent surface of the target assembly and the adjacent assembly is the same, so that the infinite grid calculation using the reflection boundary condition is a good approximation.

[0012]

However, when the D-lattice bundle and the offset bundle are adjacent to each other, the water gap widths of the adjacent surfaces are different, so that the reflection in the infinite lattice system when the respective node average nuclear constants are calculated. The approximation of the boundary condition becomes poor. This is because the thermal neutron flux distribution near the water gap changes locally due to the change in the water gap width, and the nuclear properties of the aggregate change. It should be noted that the fast neutron flux hardly changes, and the change in the nuclear properties of the aggregate is almost caused by the change in the thermal neutron flux near the water gap. This change occurs for both adjacent D grating bundles and offset bundles.

When the offset amount is large, the influence on the core characteristics of the assembly is large, and in the core performance calculation, the core critical eigenvalue, the power distribution in the reactor, the fuel rod power distribution in the assembly, the thermal limit value, the neutron detector It is necessary to consider the count value, thermal neutron flux change at the control rod position, etc. in monitoring the core performance.

As shown in FIG. 18, in the example of the nuclear characteristic change caused by the offset bundle, it is now assumed that two offset bundles having the center of the channel box offset by 2 mm toward the control rod are loaded around the control rod. FIG. 19 shows characteristics in which the local power distribution in the assembly in the D lattice bundle in this case is compared with the fuel rod local power distribution of the infinite lattice system. In FIG. 19, the upper part shows the mixed system, the middle part shows the infinite lattice system, and the lower part shows the difference.

In this case, the evaluation of the local output distribution of the reference solution was performed by a multi-aggregate heterogeneous calculation. With adjacent offset bundles, the wide gap of the D grating bundle will be substantially reduced and the narrow gap will be increased. According to the reference solution, the maximum local peaking coefficient of the aggregate of the D lattice bundle changes by about 10% due to the adjacency of the offset bundle, indicating the importance of monitoring the core in consideration of the adjacency effect with the offset bundle. Have been.

On the other hand, a similar problem occurs when the channel box 2 is deformed by fast neutron irradiation. That is, the channel box 2 is formed of, for example, a zirconium alloy. The upper and lower ends of the fuel rod are supported by an upper tie plate and a lower tie plate. The fuel assembly 1 is under conditions of high temperature, high pressure and high radiation, and the channel box 2 is deformed by fast neutron irradiation.

If there is a difference in the fast neutron irradiation amount between the opposing surfaces of the channel box 2, as shown in FIG.
Since there is a difference in the amount of irradiation growth of the channel box 2 in the axial direction, bending occurs in the axial direction. FIG. 20 shows a state in which bending occurs in the axial direction due to irradiation growth of the channel box 2 by fast neutron irradiation.

When the channel box 2 is bent as described above, the width of the water gap 4 of the fuel assembly 1 changes from the normal width as shown in FIG. Therefore, mechanically, there is an effect that the insertion of the control rod 5 into the water gap 4 is hindered.

When a channel box of the target assembly and the adjacent assembly is bent by the irradiation growth, the width of the water gap between the adjacent fuel assemblies changes from the normal width, and the nuclear characteristics of the fuel assembly are changed. Changes. Therefore, in the core performance calculation, it is important to consider the core performance change due to the channel box deformation in monitoring the core performance.

That is, when the channel box of the target assembly and the adjacent assembly bends due to the irradiation growth, the width of the water gap between the adjacent fuel assemblies changes from the normal width, and the influence on the core performance calculation is made. When the water gap width changes, the neutron moderating and absorbing effects are changed, the thermal neutron flux distribution near the water gap changes, and the nuclear characteristics of the fuel assembly change.

It should be noted that the fast neutron flux hardly changes, and the change in the nucleus characteristics of the aggregate can be said to be almost caused by the change in the thermal neutron flux near the water gap. If the channel box deformation is large, the effect on the core characteristics of the assembly is large, and in the core performance calculation, the core critical eigenvalue, reactor power distribution, fuel rod power distribution in the assembly, thermal limit value, and neutron detection due to the channel box deformation It is necessary to consider the reactor count value, thermal neutron flux change at the control rod position, etc. in monitoring the core performance.

Conventionally, the deformation of the channel box 2 due to neutron irradiation is predicted, and the control rod 5 and the channel box 2
Operational management of the core to avoid interference with
JP-A-2-176497, "Method and apparatus for evaluating channel box deformation" and JP-A-2-201
No. 291, "Operation method of a nuclear reactor"
No. 4084, “Reuse method of channel box”.

However, these documents do not consider the influence of the deformation of the channel box on the core performance calculation. An example of a nuclear property change due to channel box deformation will be described. For example, as shown in FIG. 21, the water gap 4 on the control rod 5 side of the fuel assembly 1 of interest is 2 mm due to the deformation of the channel box 2.
The water gap 4 on the instrumentation tube 6 side for neutron detector
mm, and there is no deformation in the other adjacent aggregates 1. In this case, the result of comparison between the local power distribution in the assembly and the fuel rod local power distribution of the infinite lattice system has characteristics similar to those in FIG. 19 when the offset bundle is employed.

In this case, the evaluation of the local output distribution of the reference solution was performed by a multi-aggregate heterogeneous calculation. According to this,
The maximum local peaking coefficient of the aggregate changes by about 10% due to the channel box deformation, indicating the importance of performing core monitoring in consideration of the channel box deformation.

In the calculation of the core performance, the thermal limit value is evaluated in consideration of the influence on the fuel rod output due to the deviation of the fuel assembly from the designed normal position. It is indicated by the publication "core monitoring device".

In Japanese Patent Application Laid-Open No. 4-110698, four fuel assemblies surrounding a control rod in a reactor core or a neutron detector in a reactor are defined as one unit cell, and a fuel due to misalignment of the assembly in each unit cell. It is calculated to calculate the bar output change.

In general, the change in the fuel rod output in the focused assembly in the unit cell depends on the combination of the displacement amount and the direction of the displacement of the four fuel assemblies in the unit cell.
Japanese Unexamined Patent Publication No. 4-110698 describes that the calculation is simply made from the displacement of the fuel assembly.
Evaluation method is not clear.

In Japanese Unexamined Patent Publication No. 4-110698, the influence of the displacement of the assembly adjacent to the assembly of interest from the outside of the unit cell of interest is neglected, and the change in fuel rod output is accurately evaluated. There was a problem that it was not possible.
Furthermore, from the fuel rod output change due to the misalignment of the assembly,
There is no specific method for evaluating the limit output ratio.

In Japanese Patent Application Laid-Open No. 4-110984, the critical eigenvalue of the reactor core and the power distribution in the reactor due to the misalignment of the fuel assembly, the count value of the neutron detector in the reactor and the irradiation amount of the detector, the control rod irradiation The effect on quantity etc. is not considered at all. In particular, the count of the neutron flux detector in the reactor is sensitive to changes in the water gap width.To learn the error from the measured value of the neutron flux detector and correct the in-core power distribution, the calculated value of the detector must be accurate. You need to ask for it.

Further, when the combustion proceeds with the fuel assembly being displaced, the burning history effect and the burnup distribution effect of the neutron spectrum change in the fuel assembly due to the water gap width change are accumulated, and the assembly is instantaneously positioned. There is a difference from the effect of the shift. This hysteresis effect generally occurs in a direction to offset the instantaneous effect. However, Japanese Patent Application Laid-Open No. H4-110698 does not consider this hysteresis effect at all.

Regarding the deformation of the channel box, Japanese Patent Application Laid-Open No. 4-110698 discloses that
It does not describe how to evaluate the misalignment of the aggregated channel box from the normal position due to the channel box deformation due to neutron irradiation. Therefore, it is not possible to predict the channel box deformation due to irradiation growth and calculate the thermal limit value There was a problem.

Further, when the combustion proceeds while the aggregate is displaced due to the channel box deformation, the burning history effect and the burn-up distribution effect of the neutron spectrum change in the aggregate due to the water gap width change are accumulated, and the aggregate is instantaneous. Is different from the effect when the position shifts. This hysteresis effect generally occurs in a direction to offset the instantaneous effect.
No. 0698 does not consider this hysteresis effect at all.

An object of the present invention is to provide a core performance calculation device capable of accurately calculating a core characteristic in consideration of an instantaneous and hysteresis effect when an offset bundle is adjacent to a D lattice core or when a channel box is deformed. That is.

[0034]

According to a first aspect of the present invention, there is provided a D-lattice bundle in which the water gap width on the control rod side is wider than the other side, and an offset bundle in which the channel center of the D-lattice bundle is eccentric. An apparatus for calculating core performance of a mixed boiling water reactor, wherein an average nuclear constant or an infinite lattice constant is estimated in advance by a detailed multi-aggregate calculation according to a combination of an adjacent D lattice bundle and an offset bundle. Based on a core diffusion calculation using a nuclear constant calculated by the aggregate average nuclear constant calculation means, and a critical eigenvalue of the core, A reactor power distribution calculating means for calculating a neutron flux distribution and a power distribution.

According to the first aspect of the present invention, the aggregate average nuclear constant calculating means includes an aggregate average nuclear constant or an infinite lattice estimated in advance by multi-aggregate detailed calculation in accordance with a combination of adjacent D lattice bundles and offset bundles. The aggregate average nuclear constant is calculated using the correction amount for the average nuclear constant, and the in-core power distribution calculating means uses the nuclear constant calculated by the aggregate average nuclear constant calculating means based on the core diffusion calculation, Calculate critical eigenvalue, neutron flux distribution, and power distribution.

According to a second aspect of the present invention, there is provided a boiling water atom in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is eccentric. A core performance calculation device for a reactor, based on core diffusion calculation, a critical eigenvalue of the core, a neutron flux distribution, an in-core power distribution calculating means for calculating a power distribution,
Calculate local fuel rod output using the correction amount for the fuel rod local output or infinite grid fuel rod local output in the fuel assembly that has been evaluated in advance by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and offset bundle. A local power distribution calculating means, and a line for calculating a linear power density of the assembly by using the assembly cross-sectional average power calculated by the in-furnace power distribution calculating means and the fuel rod local power calculated by the local power distribution calculating means Power density calculating means.

In the present invention, the in-core power distribution calculating means calculates a critical eigenvalue, a neutron flux distribution, and a power distribution of the core based on the core diffusion calculation, and the local power distribution calculating means calculates an adjacent D grid The fuel rod local output is calculated using a correction amount for the fuel rod local output or the infinite lattice fuel rod local output in the fuel assembly that has been evaluated in advance by the multi-assembly detailed calculation according to the combination of the bundle and the offset bundle. Then, the linear power density calculating means calculates the linear power density of the assembly using the assembly cross-sectional average power calculated by the in-furnace power distribution calculating means and the fuel rod local power calculated by the local power distribution calculating means. .

According to a third aspect of the present invention, there is provided a boiling water atom in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is decentered are mixed. A core performance calculation device for a reactor, based on core diffusion calculation, a critical eigenvalue of the core, a neutron flux distribution, an in-core power distribution calculating means for calculating a power distribution,
Calculate the fuel rod R-factor using the correction amount for the fuel rod R-factor in the fuel assembly or the infinite lattice fuel-rod R-factor which has been evaluated in advance by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and offset bundle. And
Limit power ratio calculating means for calculating a limit power ratio of the assembly using the average power of the assembly calculated by the in-core power distribution calculating means and the calculated fuel rod R factor.

According to the third aspect of the present invention, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the core diffusion calculation. The limit power ratio calculating means calculates a correction amount for the fuel rod R factor or the infinite grid fuel rod R factor in the fuel assembly previously evaluated by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and the offset bundle. Is used to calculate the fuel rod R factor, and the aggregate power limit calculated by the in-furnace power distribution calculation means and the calculated fuel rod R factor are used to calculate the limit power ratio of the fuel rod.

According to a fourth aspect of the present invention, in the second or third aspect, a multi-set for a combination of a D lattice bundle and an offset bundle having a strictest thermal limit value such as a maximum linear power density or a critical power ratio. Using the correction amount for the fuel rod local power and the fuel rod R factor previously evaluated by the body detailed calculation, the core linear power density and the critical power ratio are corrected and calculated by the core diffusion calculation. It is.

According to the fourth aspect of the present invention, in addition to the operation of the second or third aspect, the linear output density calculating means sets a maximum linear output density or a limit of a thermal limit value such as a limit output ratio. Using the correction amount for the local power of the fuel rod in the fuel assembly, which was previously evaluated by the detailed calculation of the multi-assembly for the combination of the lattice bundle and the offset bundle, the core power density was corrected by the core diffusion calculation using the core diffusion calculation, and the critical power ratio The calculating means evaluates in advance a multi-aggregate detailed calculation for the combination of the D lattice bundle and the offset bundle in which the thermal limit such as the maximum linear power density or the critical power ratio becomes the strictest, and the fuel rod local output in the fuel assembly; Using the correction amount for the fuel rod R factor, the critical power ratio is corrected and calculated by core diffusion calculation.

According to a fifth aspect of the present invention, there is provided a boiling water atom in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is decentered are mixed. A core performance calculation device for a reactor, based on core diffusion calculation, a critical eigenvalue of the core, a neutron flux distribution, an in-core power distribution calculating means for calculating a power distribution,
In-core calculated using the correction amount for the neutron flux at the in-core measuring device or the neutron flux at the infinite-lattice measuring device, which was previously evaluated by the detailed calculation of the multi-assembly according to the combination of the adjacent D lattice bundle and offset bundle A neutron flux meter in the reactor for calculating the neutron flux meter in the reactor and the irradiation amount in the reactor based on the average cross section power calculated by the neutron flux in the measuring device and the reactor power distribution calculating means; is there.

In the invention of claim 5, the in-core power distribution calculating means calculates a critical eigenvalue, a neutron flux distribution, and a power distribution of the core based on the core diffusion calculation. The in-reactor neutron flux meter counting value calculating means calculates the in-reactor in-core neutron flux or infinite lattice in-reactor position in the in-furnace measuring device, which has been evaluated in advance by multi-aggregate detailed calculation according to the combination of the adjacent D lattice bundle and offset bundle. The in-core neutron flux meter count and irradiation amount are calculated from the in-core neutron flux in the reactor, which is calculated using the correction amount for the neutron flux in the reactor, and the cross-sectional average power calculated by the in-core power distribution calculation means.

A sixth aspect of the present invention is a boiling water reactor in which a D-lattice bundle in which the water gap width of the control rod example is wider than the other side and an offset bundle in which the channel center of the D-lattice bundle is eccentric are mixed. Core performance calculation device, based on the core diffusion calculation, the critical eigenvalue of the core, neutron flux distribution, in-core power distribution calculation means for calculating the power distribution,
The control rod position neutron flux calculated using the correction amount for the control rod position neutron flux or the infinite lattice control rod position neutron flux previously evaluated by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and the offset bundle, and Control rod irradiation amount calculation means for calculating the control rod irradiation amount from the aggregate cross-sectional average power calculated by the in-furnace power distribution calculation means.

In the invention of claim 6, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the core diffusion calculation. The control rod irradiation amount calculating means calculates a correction amount for the control rod position neutron flux or the infinite lattice control rod position neutron flux evaluated in advance by the multi-aggregate detailed calculation according to the combination of the adjacent D lattice bundle and the offset bundle. The control rod irradiation amount is calculated from the control rod position neutron flux calculated using the above and the aggregate cross-sectional average power calculated by the reactor power distribution calculation means.

The invention of claim 7 is the invention of claims 2 to 5
In the invention, the aggregate average output calculated and corrected by the aggregate average nuclear constant calculating means is used as the aggregate average output.

According to a seventh aspect of the present invention, in addition to the functions of the second to fifth aspects of the present invention, an aggregate average output corrected and calculated by the first aspect of the present invention is used.

The invention of claim 8 is a boiling water atom in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is eccentric are mixed. A thermal neutron flux calculation device for a reactor core, which calculates the change of the thermal neutron flux distribution in the assembly with respect to the variation of the water gap width of the focused assembly from the infinite lattice system based on the analytical model for the diffusion equation Change calculation means, linear power density calculation means for correcting the fuel rod local output using the thermal neutron flux distribution change calculated by the thermal neutron flux change calculation means to calculate the linear power density of the assembly, and thermal neutron flux The critical power ratio calculating means for correcting the fuel rod R-factor using the thermal neutron flux distribution change calculated by the change calculating means to calculate the critical power ratio of the assembly, and the thermal neutron flux change calculating means Using the calculated thermal neutron flux distribution change to correct the neutron flux at the in-core neutron detector and calculate the in-core neutron flux meter counting and irradiation dose; Control rod irradiation amount calculating means for calculating the control rod irradiation amount by correcting the control rod position neutron flux using the thermal neutron flux distribution change calculated by the flux change calculating means.

In the invention of claim 8, the thermal neutron flux change calculating means calculates the change of the thermal neutron flux distribution in the assembly with respect to the variation of the water gap width of the assembly of interest from the infinite lattice system by an analytical model for the diffusion equation. Calculate based on The linear power density calculating means corrects the fuel rod local output using the thermal neutron flux distribution change calculated by the thermal neutron flux change calculating means to calculate the linear power density of the assembly, and the critical power ratio calculating means. Calculates the critical power ratio of the assembly by correcting the fuel rod R-factor using the thermal neutron flux distribution change calculated by the thermal neutron flux change calculating means. Further, the in-core neutron flux meter counting value calculating means corrects the in-core measuring instrument position neutron flux using the thermal neutron flux distribution change calculated by the thermal neutron flux change calculating means, and calculates the in-core neutron flux counting instrument. The control rod irradiation amount calculation means calculates the control rod irradiation amount by correcting the control rod position neutron flux using the thermal neutron flux distribution change calculated by the thermal neutron flux change calculation means.

A ninth aspect of the present invention provides an in-core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation, and a diffusion theory using a neutron flux discontinuity factor. The distribution of the homogeneous thermal neutron flux in the node where each assembly was homogenized based on the calculated values was evaluated in the infinite lattice calculation, and the "ratio of the inhomogeneous thermal neutron flux at the assembly boundary to the homogeneous thermal neutron flux" and " The ratio of the ratio of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the body is multiplied by a factor to correct the transient component due to thermal neutron diffusion of the homogeneous thermal neutron flux distribution, thereby obtaining the heterogeneous thermal neutron flux in the aggregate. Non-homogeneous thermal neutron flux calculating means for calculating the distribution,
The average nuclear constant of the aggregate is corrected using the heterogeneous neutron flux distribution by the heterogeneous thermal neutron flux calculation means to calculate the core eigenvalue, the neutron flux distribution in the reactor, and the power distribution.

According to the ninth aspect of the present invention, the in-core power distribution calculating means calculates a critical eigenvalue, a neutron flux distribution, and a power distribution of the core based on the core diffusion calculation. Then, the inhomogeneous thermal neutron flux calculation means calculates the homogeneous thermal neutron flux distribution in the node where each assembly is homogenized based on the diffusion theory using the neutron flux discontinuity factor, and was evaluated in the infinite lattice calculation. The ratio of the ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux at the aggregate boundary multiplied by the ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the aggregate is multiplied by the homogeneous thermal neutron flux distribution. Calculates the heterogeneous thermal neutron flux distribution in the assembly by using the correction factor of the transient component due to thermal neutron diffusion of, and corrects the aggregate average nuclear constant using the heterogeneous neutron flux distribution by the heterogeneous thermal neutron flux calculation means To calculate core eigenvalues, neutron flux distribution in the reactor, and power distribution.

A tenth aspect of the present invention is based on a core diffusion calculation method for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation, and a diffusion theory using a neutron flux discontinuity factor. The distribution of the homogeneous thermal neutron flux in the nodes where the aggregates were homogenized was calculated by the infinite lattice calculation, and the "ratio of the inhomogeneous thermal neutron flux at the aggregate boundary to the homogeneous thermal neutron flux" and " Multiplied by the ratio of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux at this time, is used as a correction factor for the transient component due to thermal neutron diffusion of the homogeneous thermal neutron flux distribution to obtain the heterogeneous thermal neutron flux distribution in the assembly. Neutron flux calculation means for calculating the average cross-sectional average power calculated by the in-core power distribution calculation means and the fuel calculated from the heterogeneous neutron flux distribution by the non-homogeneous thermal neutron flux calculation means Those having a linear power density calculating means for calculating a linear power density of the fuel assembly by using a local power.

In the tenth aspect of the present invention, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the core diffusion calculation. Then, the inhomogeneous neutron flux calculation means calculates a homogeneous thermal neutron flux distribution in the node where each assembly is homogenized based on the diffusion theory using a neutron flux discontinuity factor, and is evaluated in the infinite lattice calculation. The ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux at the aggregate boundary and the ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the aggregate are multiplied by the factor The non-homogeneous thermal neutron flux distribution in the assembly is calculated by using the correction factor for the transient component due to thermal neutron diffusion, and the linear power density calculation means calculates the aggregate cross-sectional average power and the non-uniform The linear power density of the fuel assembly is calculated by using the fuel rod local output calculated from the heterogeneous neutron flux distribution by the homogeneous thermal neutron flux calculation means.

The invention according to claim 11 is based on a diffusion theory using a neutron flux discontinuity factor, and a core power distribution calculation means for calculating a critical eigenvalue, a neutron flux distribution and a power distribution of the core based on the core diffusion calculation. The distribution of the homogeneous thermal neutron flux in the nodes where the aggregates were homogenized was calculated by the infinite lattice calculation, and the "ratio of the inhomogeneous thermal neutron flux at the aggregate boundary to the homogeneous thermal neutron flux" and " Multiplied by the ratio of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux at this time, is used as a correction factor for the transient component due to thermal neutron diffusion of the homogeneous thermal neutron flux distribution to obtain the heterogeneous thermal neutron flux distribution in the assembly. Neutron flux calculation means for calculating the average cross-sectional average power calculated by the in-core power distribution calculation means and the fuel calculated from the heterogeneous neutron flux distribution by the non-homogeneous thermal neutron flux calculation means A linear power density calculation means for calculating a linear power density of the fuel assembly by using a local power,
A critical power ratio for calculating a critical power ratio of a fuel assembly from a fuel rod R factor calculated using the average cross-sectional power of the assembly calculated by the reactor power distribution calculating means and the fuel rod local output by the linear power density calculating means. Calculation means.

In the eleventh aspect of the present invention, the in-core power distribution calculating means calculates a critical eigenvalue, a neutron flux distribution, and a power distribution of the core based on the core diffusion calculation. Then, the inhomogeneous neutron flux calculation means calculates a homogeneous thermal neutron flux distribution in the node where each assembly is homogenized based on the diffusion theory using a neutron flux discontinuity factor, and is evaluated in the infinite lattice calculation. The ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux at the aggregate boundary and the ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the aggregate are multiplied by the factor The inhomogeneous thermal neutron flux distribution in the assembly is calculated by using the correction factor for the transient component due to thermal neutron diffusion. In addition, the linear power density calculating means uses the fuel rod local output calculated from the aggregate cross-sectional average power calculated by the in-core power distribution calculating means and the inhomogeneous neutron flux distribution by the inhomogeneous thermal neutron flux calculating means. The linear power density of the fuel assembly is calculated, and the critical power ratio calculating means is calculated using the average output of the cross section of the assembly calculated by the in-core power distribution calculating means and the local power of the fuel rod by the linear power density calculating means. The critical power ratio of the fuel assembly is calculated from the fuel rod R factor.

According to a twelfth aspect of the present invention, there is provided a core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution and a power distribution of a core based on a core diffusion calculation, and a diffusion theory calculation using a neutron flux discontinuity factor. A node boundary value calculating means for calculating a neutron flux and a neutron flow at a node boundary surface in which each assembly is homogenized, and an aggregate section average output and a node boundary value calculating means calculated by the in-core power distribution calculating means Linear power density calculating means for calculating the linear power density of the assembly by the infinite lattice fuel rod local power corrected using the sensitivity coefficient of the fuel rod local power for neutron flux and neutron flow on each surface of the assembly boundary It is.

In the twelfth aspect, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the core diffusion calculation. Then, the node boundary value calculating means calculates a neutron flux and a neutron flux at the node boundary surface where the respective assemblies are homogenized based on a diffusion theory calculation using a neutron flux discontinuity factor, and the linear output density calculating means ,
Infinite grid fuel rods corrected by using the cross-sectional average power calculated by the in-core power distribution calculation means and the sensitivity coefficient of the local power of the fuel rods to neutron flux and neutron flow on each side of the assembly boundary calculated by the node boundary value calculation means Calculate the linear output density of the aggregate by local output.

According to a thirteenth aspect of the present invention, there is provided a core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation, and a diffusion theory calculation using a neutron flux discontinuity factor. A node boundary value calculating means for calculating a neutron flux and a neutron flow at a node boundary surface in which each assembly is homogenized, and an aggregate section average output and a node boundary value calculating means calculated by the in-core power distribution calculating means A linear power density calculating means for calculating a linear power density of an assembly by an infinite grid fuel rod local output corrected using a sensitivity coefficient of a fuel rod local output for neutron flux and neutron flow on each surface of an assembly boundary; The limit of the fuel assembly is obtained from the fuel rod R factor calculated using the average output of the cross section of the assembly calculated by the power distribution calculating means and the local output of the fuel rod by the linear power density calculating means. Those having a critical power ratio calculating means for calculating the ratio.

According to a thirteenth aspect of the present invention, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the core diffusion calculation. Then, the linear output density calculating means is based on a diffusion theory calculation using a neutron flux discontinuity factor, and a node boundary value calculating means for calculating a neutron flux and a neutron flow of a node boundary surface in which each assembly is homogenized,
Infinite grid fuel rods corrected by using the cross-sectional average power calculated by the in-core power distribution calculation means and the sensitivity coefficient of the local power of the fuel rods to neutron flux and neutron flow on each side of the assembly boundary calculated by the node boundary value calculation means Calculate the linear output density of the aggregate by local output. Further, the limit power ratio calculating means calculates the fuel assembly from the fuel rod R-factor calculated by using the fuel rod local output and the fuel rod local output calculated by the reactor power distribution calculating means. Calculate the limit power ratio.

According to a fourteenth aspect of the present invention, there is provided a core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation, and a diffusion theory calculation using a neutron flux discontinuity factor. A node boundary value calculating means for calculating a neutron flux and a neutron flux at a node boundary surface obtained by homogenizing each of the assemblies, and an aggregate average output and a node boundary value calculating means calculated by the in-core power distribution calculating means. A limit power ratio calculating means for calculating a limit power ratio of the assembly by an infinite lattice R factor corrected using a sensitivity factor of a fuel rod R factor for neutron flux and neutron flow on each surface of the assembly boundary.

According to the invention of claim 14, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution and power distribution of the core based on the core diffusion calculation, and the node boundary value calculating means calculates the neutron flux discontinuity. The neutron flux and neutron flow are calculated based on the diffusion theory calculation using the factors. The limit power ratio calculation means calculates the average coefficient of the fuel rod R and the neutron flux and the neutron flow of the fuel rod R factor for each surface of the assembly boundary calculated by the in-core power distribution calculation means and the node boundary value calculation means. The limit power ratio of the aggregate is calculated by the infinite lattice R factor corrected using the above.

According to a fifteenth aspect of the present invention, there is provided an in-core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution and a power distribution of a core based on a core diffusion calculation, and a diffusion theory calculation using a neutron flux discontinuity factor. A node boundary value calculating means for calculating a neutron flux and a neutron flow at a node boundary surface in which each assembly is homogenized, and an aggregate section average output and a node boundary value calculating means calculated by the in-core power distribution calculating means Reactor that calculates the neutron flux meter count and irradiation dose from the neutron flux in the infinite grid detector corrected using the neutron flux at the in-furnace detector in the reactor for neutron flux and neutron flow on each surface of the assembly boundary And an internal neutron flux meter counting value calculating means.

According to the fifteenth aspect, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the core diffusion calculation, and the node boundary value calculating means calculates the neutron flux discontinuity. The neutron flux and neutron flow are calculated based on the diffusion theory calculation using the factors. The in-reactor neutron flux meter counting value calculating means includes an in-reactor for the neutron flux and neutron flow on each surface of the aggregate boundary calculated by the in-core power distribution calculating means and the aggregate boundary by the node boundary value calculating means. From the neutron flux at the infinite grid detector corrected by using the sensitivity coefficient of the neutron flux at the measuring instrument, the neutron flux counting in the reactor and the irradiation dose are calculated.

A sixteenth aspect of the present invention provides a core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation, and a diffusion theory calculation using a neutron flux discontinuity factor. A node boundary value calculating means for calculating a neutron flux and a neutron flow at a node boundary surface in which each assembly is homogenized, and an aggregate section average output and a node boundary value calculating means calculated by the in-core power distribution calculating means Control rod irradiation amount calculation means for calculating the control rod irradiation amount from the neutron flux at the infinite grid control rod position corrected using the neutron flux sensitivity coefficient of the control rod position for neutron flux and neutron flow on each surface of the aggregate boundary Things.

In the invention of claim 16, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution and power distribution of the core based on the core diffusion calculation, and the node boundary value calculating means calculates the neutron flux discontinuity. The neutron flux and neutron flow are calculated based on the diffusion theory calculation using the factors. Then, the control rod irradiation amount calculation means includes a neutron flux of the control rod position neutron flux and a neutron flux and a neutron flux on each surface of the aggregate boundary calculated by the in-furnace power distribution calculation means and the node boundary value calculation means. The control rod dose is calculated from the neutron flux at the infinite lattice control rod position corrected using the sensitivity coefficient.

According to a seventeenth aspect of the present invention, in the ninth to sixteenth aspects of the present invention, the in-furnace power distribution calculating means includes a first modification.
Calculate the fast neutron flux distribution by group diffusion calculation and calculate the thermal neutron flux and thermal neutron flow using the obtained fast neutron flux and the spectral index that is the ratio of the thermal neutron flux to the fast neutron flux in the infinite lattice calculation. It is like that.

According to the seventeenth aspect, in addition to the effects of the ninth to sixteenth aspects, the fast neutron flux distribution is calculated by the modified one-group diffusion calculation, and the obtained fast neutron flux and the heat in the infinite lattice calculation are calculated. The thermal neutron flux and the thermal neutron flux are calculated using the spectral index, which is the ratio of the neutron flux to the fast neutron flux.

According to an eighteenth aspect of the present invention, in the first to seventeenth aspects of the present invention, the calculated value of the in-core neutron flux measurement device in which the adjacent effect between the D lattice bundle and the offset bundle is corrected is used as the in-core neutron. An in-furnace power correction means for correcting the in-furnace power distribution calculation value by adapting the measured value of the bundle measuring device is provided.

According to the eighteenth aspect of the present invention, in addition to the effects of the first to seventeenth aspects, the calculated value of the in-core neutron flux measuring instrument count value in which the adjacent effect between the D lattice bundle and the offset bundle is corrected is Correct the calculated power distribution in the reactor according to the measured value of the neutron flux detector in the reactor.

A nineteenth aspect of the present invention is the invention according to the ninth to eighteenth aspects, wherein the burn-up distribution in the fuel assembly and the spectrum history distribution which is the burn-up average value of the ratio of the thermal neutron flux to the fast neutron flux are used. Thus, the correction amount of the aggregate nuclear constant, the fuel rod local power distribution, and the fuel rod R-factor distribution due to the hysteresis effect associated with the adjacency between the D lattice bundle and the offset bundle is calculated.

According to the nineteenth aspect of the present invention, in addition to the effects of the ninth to eighteenth aspects, the burn-up distribution in the fuel assembly and the spectrum history which is the burn-up average value of the ratio of the thermal neutron flux to the fast neutron flux are provided. The distribution is used to calculate the correction amount of the aggregate nuclear constant, the fuel rod local power distribution, and the fuel rod R-factor distribution due to the hysteresis effect associated with the adjacency between the D lattice bundle and the offset bundle.

A twentieth aspect of the present invention provides a reactor power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a channel box dose calculating means for calculating a fast neutron dose on each surface of a channel box, Channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means, and channel box deformation amount calculating means Calculate the aggregate nuclear constant and the neutron flux discontinuity factor with the channel box deformation effect corrected using the channel box deformation amount evaluated by, and use the corrected aggregate nuclear constant and the neutron flux discontinuity factor to calculate the core With a core power distribution recalculation means for calculating a critical eigenvalue and a power distribution. That.

In the twentieth aspect, the in-core power distribution calculating means calculates the neutron flux and the power distribution in the core, and the channel box dose calculating means calculates the fast neutron dose on each surface of the channel box. The channel box deformation amount calculating means evaluates the deformation amount of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means. Then, the in-reactor power distribution recalculating means calculates the aggregate nuclear constant and the neutron flux discontinuity factor in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculating means, and is corrected. The critical eigenvalue and power distribution of the reactor core are calculated using the aggregate nuclear constant and the neutron flux discontinuity factor.

A twenty-first aspect of the present invention provides a reactor power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a channel box dose calculating means for calculating a fast neutron dose on each surface of the channel box, Channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means, and channel box deformation amount calculating means Calculates the local power of the fuel rods in the fuel assembly in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the above, and gives the maximum local power peaking coefficient of the assembly and the set obtained by the core power distribution calculation means. Using the average power of the cross section of the fuel, the linear output of the cross section of each fuel assembly Those having a linear power density calculation means for calculating degrees.

According to the twenty-first aspect of the present invention, the neutron flux and the power distribution in the reactor core are calculated by the in-core power distribution calculating means, and the fast neutron irradiation amount on each surface of the channel box is calculated by the channel box irradiation amount calculating means. The channel box deformation amount calculating means evaluates the deformation amount of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means. Then, the linear output density calculation means calculates the fuel rod local output in the fuel assembly in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculation means, Is calculated using the maximum local power peaking coefficient and the average power of the cross section of the assembly by the in-core power distribution calculation means.

A twenty-second aspect of the present invention provides a reactor power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a channel box dose calculating means for calculating a fast neutron dose on each surface of the channel box, Channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means, and channel box deformation amount calculating means Calculates the fuel rod R factor representing the fuel rod output pattern in the assembly, in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the above, and calculates the maximum R factor of the assembly and the set by the core power distribution calculation means. Using the average fuel output, the monitoring index of burnout of the fuel rods of each fuel assembly That is one having a critical power ratio calculating means for calculating a critical power ratio.

According to the invention of claim 22, the in-core power distribution calculating means calculates the neutron flux and the power distribution in the core, and the channel box irradiation amount calculating means calculates the fast neutron irradiation amount on each surface of the channel box. The channel box deformation amount calculating means evaluates the deformation amount of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means. Then, the limit power ratio calculating means calculates a fuel rod R factor representing a fuel rod output pattern in the assembly in which the channel box deformation effect has been corrected using the channel box deformation amount evaluated by the channel box deformation amount calculating means. Using the maximum R factor of the fuel assembly and the average power of the fuel assembly by the core power distribution calculating means, a critical power ratio, which is a monitoring index of burnout of a fuel rod of each fuel assembly, is calculated.

A twenty-third aspect of the present invention provides a reactor power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a channel box dose calculating means for calculating a fast neutron dose on each surface of a channel box, Channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means, and channel box deformation amount calculating means Neutron flux meter measurement value calculating means for calculating the in-core neutron flux meter count value and the in-core neutron flux meter irradiation amount in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by Things.

According to a twenty-third aspect of the present invention, the neutron flux and the power distribution in the reactor core are calculated by the in-core power distribution calculating means, and the fast neutron irradiation amount on each surface of the channel box is calculated by the channel box irradiation amount calculating means. The channel box deformation amount calculating means evaluates the deformation amount of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means. The in-reactor neutron flux measurement device calculation value calculation means includes an in-core neutron flux measurement device count value and an in-core neutron flux correction device for correcting the channel box deformation effect using the channel box deformation amount evaluated by the channel box deformation amount calculation means. Calculate the meter exposure.

According to a twenty-fourth aspect of the present invention, there is provided a reactor power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a channel box dose calculating means for calculating a fast neutron dose on each surface of the channel box, Channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means, and channel box deformation amount calculating means Control rod irradiation amount calculating means for calculating the control rod irradiation amount using the control rod position neutron flux in which the channel box deformation effect is corrected by the channel box deformation amount evaluated by (1).

In the invention of claim 24, the neutron flux and the power distribution in the reactor core are calculated by the reactor power distribution calculating means, and the fast neutron irradiation amount on each surface of the channel box is calculated by the channel box irradiation amount calculating means. The channel box deformation amount calculating means evaluates the deformation amount of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means. The control rod irradiation amount calculating means calculates the control rod irradiation amount using the control rod position neutron flux in which the channel box deformation effect is corrected by the channel box deformation amount evaluated by the channel box deformation amount calculating means.

According to a twenty-fifth aspect of the present invention, in the twentieth to twenty-fourth aspects, the channel box deformation amount calculating means calculates the irradiation growth amount on the opposite surface of the channel box as a function of the fast neutron irradiation amount on each surface. Calculate,
The amount of deformation of the channel box is calculated from the difference in the amount of irradiation growth on the opposing surface.

According to the twenty-fifth aspect of the present invention, in addition to the functions of the twentieth to twenty-fourth aspects, when the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation model, The amount of irradiation growth is calculated as a function of the amount of fast neutron irradiation on each surface, and the amount of deformation of the channel box is calculated from the difference between the amounts of irradiation growth on the opposing surfaces.

According to a twenty-sixth aspect of the present invention, there is provided an in-core power calculating means for calculating a neutron flux and a power distribution in a reactor core, a burn-up distribution calculating means for calculating a burn-up distribution in the reactor core, and a burn-up distribution calculating means. Using the burnup of each surface of the evaluated channel box, the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model using that the irradiation growth of the channel box corresponds one-to-one to the burnup. A channel box deformation amount calculating means, and an aggregate in which a channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculating means, a nuclear constant and a neutron flux discontinuity factor are calculated, and the corrected aggregate is calculated. Reactor power distribution is calculated by using the nuclear constant and neutron flux discontinuity factor to calculate the critical eigenvalue and power distribution of the core. It is obtained and a calculation unit.

According to the twenty-sixth aspect of the present invention, the in-core power calculation means calculates the neutron flux and power distribution in the core, the burn-up distribution calculation means calculates the burn-up distribution in the core, and the channel box deformation amount calculation means. Using the burnup of each surface of the channel box evaluated by the burnup distribution calculation means,
The amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model using the fact that the irradiation growth amount of the channel box corresponds one-to-one with the burnup. Then, the in-reactor power distribution recalculating means calculates the aggregate nuclear constant and the neutron flux discontinuity factor in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculating means, and is corrected. The core eigenvalue and power distribution of the core are calculated using the aggregate nuclear constant and the neutron flux discontinuity factor.

The invention according to claim 27 is an in-core power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a burn-up distribution calculating means for calculating a burn-up distribution in the core, and a burn-up distribution calculating means. Using the burn-up of each surface of the channel box evaluated by the above, the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model that uses that the irradiation growth of the channel box corresponds to the burn-up one-to-one Channel box deformation amount calculating means,
The local power of the fuel rods in the fuel assembly, in which the channel box deformation effect is corrected by using the channel box deformation amount evaluated by the channel box deformation amount calculating means, is calculated, and the maximum local power peaking coefficient of the assembly and the A linear power density calculating means for calculating a linear power density of each fuel assembly cross section by using the assembly cross section average output by the power distribution calculating means.

According to a twenty-seventh aspect of the present invention, the in-core power distribution calculating means calculates the neutron flux and the power distribution in the core, and the burn-up distribution calculating means calculates the burn-up distribution in the core. The means uses the burn-up of each surface of the channel box evaluated by the burn-up distribution calculation means to calculate the amount of deformation of the channel box due to neutron irradiation,
The evaluation is performed using an irradiation growth model using the fact that the irradiation growth amount of the channel box corresponds to the burnup on a one-to-one basis.
Then, the linear output density calculation means calculates the fuel rod local output in the fuel assembly in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculation means, Is calculated using the maximum local power peaking coefficient and the average power of the cross section of the assembly by the in-core power distribution calculation means.

The invention according to claim 28 is an in-core power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a burn-up distribution calculating means for calculating a burn-up distribution in the core, and a burn-up distribution calculating means. Using the burn-up of each surface of the channel box evaluated by the above, the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model that uses that the irradiation growth of the channel box corresponds to the burn-up one-to-one Channel box deformation amount calculating means,
Using the channel box deformation amount evaluated by the channel box deformation amount calculation means, the fuel rod R factor representing the fuel rod output pattern in the assembly in which the channel box deformation effect is corrected is calculated, whereby the assembly maximum R factor and the in-core Limit power ratio calculating means for calculating a limit power ratio, which is a monitoring index of burnout of fuel rods of each fuel assembly, using the average power of the assembly by the power distribution calculating means.

According to the twenty-eighth aspect of the present invention, the in-core power distribution calculating means calculates the neutron flux and the power distribution in the core, the burn-up distribution calculating means calculates the burn-up distribution in the core, and calculates the channel box deformation. The means uses the burn-up of each surface of the channel box evaluated by the burn-up distribution calculation means to calculate the amount of deformation of the channel box due to neutron irradiation,
The evaluation is performed using an irradiation growth model using the fact that the irradiation growth amount of the channel box corresponds to the burnup on a one-to-one basis.
Then, the limit power ratio calculating means calculates a fuel rod R factor representing a fuel rod output pattern in the assembly in which the channel box deformation effect is corrected by using the channel box deformation amount evaluated by the channel box deformation amount calculating means,
Thus, the critical power ratio, which is a monitoring index of burnout of the fuel rods of each fuel assembly, is calculated using the maximum R factor of the assembly and the average power of the assembly by the in-core power distribution calculation means.

A twenty-ninth aspect of the present invention provides an in-core power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a burn-up distribution calculating means for calculating a burn-up distribution in the core, and a burn-up distribution calculating means. Using the burn-up of each surface of the channel box evaluated by the above, the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model that uses that the irradiation growth of the channel box corresponds to the burn-up one-to-one Channel box deformation amount calculating means,
In-core neutron flux measurement device that calculates the in-core neutron flux measurement device count and in-core neutron flux measurement device irradiation amount that corrects the channel box deformation effect using the channel box deformation amount evaluated by the channel box deformation amount calculation means And a counting value calculating means.

According to a twenty-ninth aspect of the present invention, the neutron flux and the power distribution in the core are calculated by the in-core power distribution calculating means, and the burn-up distribution in the core is calculated by the burn-up distribution calculating means. The means uses the burn-up of each surface of the channel box evaluated by the burn-up distribution calculation means to calculate the amount of deformation of the channel box due to neutron irradiation,
The evaluation is performed using an irradiation growth model using the fact that the irradiation growth amount of the channel box corresponds to the burnup on a one-to-one basis.
The in-core neutron flux meter counting value calculating means uses the channel box deformation amount evaluated by the channel box deformation amount calculating means to correct the channel box deformation effect in-core neutron flux meter counting value and in-core neutron. Calculate the dose of the flux meter.

A thirtieth aspect of the present invention provides an in-core power distribution calculating means for calculating a neutron flux and a power distribution in a reactor core, a burn-up distribution calculating means for calculating a burn-up distribution in the core, and a burn-up distribution calculating means. Using the burn-up of each surface of the channel box evaluated by the above, the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model that uses that the irradiation growth of the channel box corresponds to the burn-up one-to-one Channel box deformation amount calculating means,
Control rod irradiation amount calculating means for calculating the control rod irradiation amount using the control rod position neutron flux in which the channel box deformation effect is corrected by the channel box deformation amount evaluated by the channel box deformation amount calculating means.

According to a thirty-first aspect of the present invention, the in-core power distribution calculating means calculates the neutron flux and the power distribution in the core, and the burn-up distribution calculating means calculates the burn-up distribution in the core. The means uses the burn-up of each surface of the channel box evaluated by the burn-up distribution calculation means to calculate the amount of deformation of the channel box due to neutron irradiation,
The evaluation is performed using an irradiation growth model using the fact that the irradiation growth amount of the channel box corresponds to the burnup on a one-to-one basis.
The control rod irradiation amount calculating means calculates the control rod irradiation amount using the control rod position neutron flux in which the channel box deformation effect is corrected by the channel box deformation amount evaluated by the channel box deformation amount calculating means.

According to a thirty-first aspect of the present invention, in the twenty-seventh to thirty-seventh aspects, the channel box deformation amount calculating means calculates the irradiation growth amount on the opposite surface of the channel box as a function of the burnup of each surface of the channel box. In this case, the amount of deformation of the channel box is calculated from the difference between the irradiation growth amounts on the opposing surfaces.

According to the thirty-first aspect, in addition to the effects of the twenty-seventh to thirty-third aspects, when the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation model, the amount of The amount of irradiation growth is calculated as a function of the burnup of each surface of the channel box, and the amount of deformation of the channel box is calculated from the difference of the amount of irradiation growth on the opposite surface.

According to a thirty-second aspect of the present invention, in the twentieth to thirty-first aspects of the present invention, the in-furnace power distribution recalculating means includes a channel box modification of the target aggregate and the adjacent aggregate evaluated in advance by the multi-aggregate detailed calculation. Average nuclear constant for the mass, fuel rod local power distribution, fuel rod maximum local power peaking coefficient, fuel rod R-factor distribution, fuel rod maximum R-factor, neutron flux at reactor neutron flux measuring instrument, neutron flux at control rod location, etc. By arranging the fuel assembly nuclear characteristics change as a table or fitting coefficient for the water gap width of each surface of the channel box of the target assembly, and referring to the water gap width obtained from the channel box deformation, This is to calculate the nuclear characteristics of the fuel assembly.

According to a thirty-second aspect of the present invention, in addition to the effects of the twentieth to thirty-first aspects, when calculating the fuel assembly core characteristics in which the channel box deformation effect is corrected, the multi-assembly detailed calculation is performed in advance. Average nuclear constant of the target aggregate and the adjacent aggregate with respect to the channel box deformation, fuel rod local power distribution, fuel rod maximum local power peaking coefficient, fuel rod R factor distribution, fuel rod maximum R factor, neutron flux in the reactor Changes in nuclear characteristics of the fuel assembly, such as the neutron flux at the measuring instrument and the neutron flux at the control rod, are arranged as a table or fitting coefficient for the water gap width of each surface of the channel box of the focused assembly, and are calculated from the channel box deformation. The nuclear characteristics of the fuel assembly are calculated by referring to the water gap width.

According to a thirty-third aspect of the present invention, in the invention of the twentieth to thirty-first aspects, the in-furnace power distribution recalculating means is configured so that the neutron flux in the aggregate with respect to the channel box deformation estimated in advance by the single aggregate detailed calculation. The change in the continuous factor
A table or fitting coefficient for the water gap width of each surface of the channel box of the target aggregate is organized and referenced using the water gap width obtained from the amount of channel box deformation. Calculate the flux distribution and use this neutron flux distribution to calculate the average nuclear constant of the aggregate, the local power distribution of the fuel rod, the maximum local power peaking coefficient of the fuel rod, the fuel rod R-factor distribution, the fuel rod maximum R-factor, and the neutron flux measurement in the reactor The nuclear characteristics of the fuel assembly such as the neutron flux at the reactor and the neutron flux at the control rod are calculated.

According to the thirty-third aspect, in addition to the effects of the twentieth to thirty-first aspects, when calculating the nuclear characteristics of the fuel assembly in which the channel box deformation effect is corrected, the detailed calculation of the single assembly is performed in advance. The change of the aggregate neutron flux discontinuity factor with respect to the evaluated channel box deformation is organized as a table or fitting coefficient for the water gap width of each surface of the channel box of the target assembly, and the water gap calculated from the channel box deformation By referring to the width, the neutron flux distribution in the assembly is calculated based on the diffusion model, and using this neutron flux distribution, the average nuclear constant of the assembly, the fuel rod local power distribution, the fuel rod maximum local power peaking coefficient , Fuel rod R-factor distribution, fuel rod maximum R-factor, neutron flux at reactor neutron flux measuring instrument, neutron flux at control rod location, etc. Calculating a fuel assembly neutronic characteristics.

According to a thirty-fourth aspect of the present invention, in the invention of the twentieth to thirty-first aspects, the in-furnace power distribution recalculating means is configured so that the neutron flux of the aggregate neutron flux with respect to the channel box deformation estimated in advance by the single aggregate detailed calculation. The change in the continuous factor
By organizing as a table or fitting coefficient for the water gap width of each surface of the channel box of the target aggregate, and referring to the water gap width obtained from the amount of channel box deformation, each boundary of the aggregate based on the diffusion model Calculate the neutron flux and neutron flow of the surface,
Aggregate average nuclear constant for neutron flux and neutron flow of each surface evaluated in advance, fuel rod local power distribution, fuel rod maximum local power peaking coefficient, fuel-like R factor distribution, fuel rod maximum R factor, reactor neutron flux measurement Using the sensitivity coefficients of the nuclear characteristics of the fuel assembly such as the neutron flux at the reactor and the neutron flux at the control rod,
This is to calculate the nuclear characteristics of the fuel assembly.

According to a thirty-fourth aspect of the present invention, in addition to the effects of the twentieth to thirty-first aspects, when calculating the fuel assembly nuclear characteristics in which the channel box deformation effect is corrected, a single assembly detailed calculation is performed in advance. The change of the aggregate neutron flux discontinuity factor with respect to the evaluated channel box deformation is organized as a table or fitting coefficient for the water gap width of each surface of the channel box of the target assembly, and the water gap calculated from the channel box deformation By referring to the width, the neutron flux and neutron flow on each surface of the assembly boundary are calculated based on the diffusion model, and the average nuclear constant for the neutron flux and neutron flow on each surface evaluated in advance, Local power distribution, fuel rod maximum local power peaking coefficient, fuel-like R-factor distribution, fuel rod maximum R-factor, furnace neutral Flux measuring instrument position neutron flux, using the response factor of the fuel assembly neutronic characteristics such as the control rod position neutron flux, to calculate the nuclear properties of the fuel assembly.

According to a thirty-fifth aspect of the present invention, in the twentieth to thirty-first aspects of the present invention, the in-furnace power distribution recalculating means includes a thermal neutron flux distribution in the aggregate with respect to the channel box deformation of the aggregate of interest and the adjacent aggregate. Is calculated based on an analytical model for the diffusion equation, and using this thermal neutron flux distribution, the aggregate average nuclear constant, fuel rod local power distribution, fuel rod maximum local power peaking coefficient, fuel rod R-factor distribution, fuel rod The nuclear characteristics of the fuel assembly, such as the rod maximum R factor, the neutron flux at the reactor neutron flux measuring device, and the neutron flux at the control rod, are calculated.

According to a thirty-fifth aspect of the present invention, in addition to the effects of the twentieth to thirty-first aspects, when calculating the fuel assembly core characteristics in which the channel box deformation effect is corrected, the target assembly and the adjacent assembly are calculated. The change of the thermal neutron flux distribution in the assembly with respect to the channel box deformation is calculated based on the analytical model for the diffusion equation, and using this thermal neutron flux distribution, the aggregate average nuclear constant, the fuel rod local power distribution,
Calculate the nuclear characteristics of the fuel assembly, such as the maximum local power peaking coefficient of the fuel rod, the distribution of the fuel rod R-factor, the maximum fuel rod R-factor, the neutron flux at the reactor neutron flux measuring device, and the neutron flux at the control rod.

According to a thirty-sixth aspect of the present invention, in the twentieth to thirty-first aspects, the calculated value of the in-core neutron flux measuring instrument in which the deformation effect of the channel box due to neutron irradiation has been corrected is calculated. The calculated value of the power distribution in the furnace is corrected by adapting the measured value of the vessel.

According to a thirty-sixth aspect of the present invention, in addition to the effects of the twentieth to thirty-first aspects, the calculated value of the in-furnace neutron flux meter counting value in which the deformation effect of the channel box due to neutron irradiation is corrected is calculated. The calculated power distribution in the reactor is corrected according to the measured value of the neutron flux measuring device.

[0106] The invention of claim 37 is the invention according to claims 21 to 24, wherein the fuel set used for calculating the linear power density, the critical power ratio, the in-reactor neutron flux meter counting and irradiation amount, and the control rod irradiation amount. The fuel assembly average output in which the channel box deformation effect is corrected according to claim 20 is used as the body average output.

According to a thirty-seventh aspect, in addition to the effects of the twenty-first to twenty-fourth aspects, the linear power density, the limit power ratio, the in-reactor neutron flux meter counting and irradiation amount, and the control rod irradiation amount are calculated. The average output of the fuel assembly used is defined in claim 2
The average output of the fuel assembly in which the channel box deformation effect is corrected by 0 is used.

[0108] The invention of claim 38 is the fuel assembly according to claims 27 to 30, which is used for calculating the linear power density, the limit power ratio, the neutron flux meter in the reactor and the irradiation amount, and the control rod irradiation amount. The fuel assembly average output obtained by correcting the channel box deformation effect according to claim 26 is used as the body average output.

According to the thirty-eighth aspect of the present invention, in addition to the effects of the twenty-seventh to thirty-third aspects, the linear power density, the limit power ratio, the in-reactor neutron flux meter counting and irradiation amount, and the control rod irradiation amount are calculated. The average output of the fuel assembly used is defined in claim 2
6, the average output of the fuel assembly corrected for the channel box deformation effect is used.

According to a thirty-ninth aspect of the present invention, in the twentieth to twenty-second or thirty-six to thirty-eighth aspects, the burn-up distribution and high-speed thermal neutron flux in the fuel assembly in consideration of the deformation effect of the channel box are provided. Using the spectral history distribution, which is the average burnup ratio of the neutron flux to the neutron flux, calculate the correction amount of the aggregate nuclear constant, fuel rod local power distribution, and fuel rod R-factor distribution due to the hysteresis effect accompanying the combustion of the channel box deformation. It is like that.

According to a thirty-ninth aspect, in addition to the effects of the twentieth to twenty-second or thirty-six to thirty-eighth aspects, the burnup distribution and thermal neutrons in the fuel assembly in consideration of the deformation effect of the channel box are provided. Using the spectral history distribution, which is the average burnup value of the ratio of the flux to the fast neutron flux, the correction amount of the aggregate nuclear constant, fuel rod local power distribution, and fuel rod R-factor distribution due to the hysteresis effect accompanying the combustion of the channel box deformation Is calculated.

According to a fortieth aspect of the present invention, in the thirty-third or thirty-fourth aspect, the fast neutron flux distribution is calculated by the modified one-group diffusion calculation, and the obtained fast neutron flux and the thermal neutron flux in the infinite lattice system are calculated. 3. The core performance calculation apparatus according to claim 1, wherein the thermal neutron flux and the thermal neutron flow are calculated using a spectrum index which is a ratio to a fast neutron flux.

According to a forty-ninth aspect, in addition to the effects of the thirty-third and thirty-fourth aspects, a fast neutron flux distribution is calculated by a modified one-group diffusion calculation, and the obtained fast neutron flux and heat in an infinite lattice system are calculated. The thermal neutron flux and the thermal neutron flux are calculated using the spectral index, which is the ratio of the neutron flux to the fast neutron flux.

According to a forty-first aspect of the present invention, when the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model, the ratio of neutrons of 1 MeV or more to the total fast neutrons is evaluated. Is calculated as a table of the burnup and the void ratio or a fitting coefficient.

According to the invention of claim 41, in addition to the effect of the invention of claim 25, when the amount of deformation of the channel box due to neutron irradiation is evaluated using an irradiation growth model, all the fast neutrons of neutrons having a neutron energy of 1 MeV or more are used. Is calculated as a table of burnup and void fraction or a fitting coefficient.

According to a forty-second aspect of the present invention, in the twenty-fifth or thirty-first aspect, the channel box deformation calculating means predicts and calculates the channel box deformation using the past measured values of the channel box deformation. It is something to do.

According to the invention of claim 42, in addition to the effect of the invention of claim 25 or 31, when the deformation amount of the channel box due to neutron irradiation is evaluated using the irradiation growth model, the past channel box deformation amount is evaluated. The channel box deformation amount is predicted and calculated using the actual measurement value of.

According to a forty-third aspect of the present invention, in the twenty-fifth or thirty-first aspect, the amount of deformation of the channel box caused by neutron irradiation is changed by the difference in the amount of irradiation growth on the opposing surface of the channel box. The bending in the direction is determined analytically as having a uniform curvature.

According to the invention of claim 43, in addition to the effect of the invention of claim 25 or 31, the amount of deformation of the channel box due to neutron irradiation is increased by the difference in the amount of irradiation growth on the opposing surface of the channel box. It is determined analytically assuming that the axial bending of the box has a uniform curvature.

According to a forty-fourth aspect of the present invention, in the twenty-fifth or thirty-first aspect, the amount of deformation of the channel box due to neutron irradiation is determined by dividing the channel box into axial nodes, and for each node, After the lengths after irradiation growth are obtained and stacked, the distance between the straight line connecting the upper and lower ends of the channels and each node is obtained as a bending amount.

According to a forty-fourth aspect, in addition to the effects of the twenty-fifth or thirty-first aspect, the amount of deformation of the channel box due to neutron irradiation is divided into the axial direction nodes of the channel box. After the lengths of the facing surfaces after irradiation growth are obtained and stacked, the distance between a straight line connecting the upper and lower ends of the channel and each node is obtained as a bending amount.

[0122]

Embodiments of the present invention will be described below. First, in the D-lattice core, the core nuclear characteristics such as the average nuclear constant of the aggregate in consideration of the offset bundle adjacent effect, the local power of the fuel rod, the fuel rod R factor, the neutron detector count in the reactor, and the control rod irradiation amount are accurately calculated. Will be described. That is, this is achieved by the following processes [1], [2], and [3].

[1] The change in the nuclear properties of the aggregate due to the adjacency of the D lattice bundle and the offset bundle is determined based on the multi-aggregate detailed calculation model or the neutron diffusion model as shown in the following (a) to (d). To calculate.

(A) Multi-aggregate detailed two-dimensional calculation model Multi-aggregate details 2 for the focused aggregate and the adjacent aggregate
By the dimension calculation, the aggregate nuclear characteristic change amount such as the aggregate average nuclear constant, the fuel rod output distribution, and the fuel rod R-factor distribution due to the proximity of the D lattice bundle and the offset bundle in advance is directly adjacent to the target aggregate. Prepare a table of combinations of aggregates with different water gap widths.

(B) Analytical diffusion model Based on the analytical solution of the two-region diffusion model in which the water gap and the fuel region were homogenized, the change in the thermal neutron flux distribution in the assembly was determined, and the substantial change in the water gap width was obtained. The change in nuclear properties is evaluated from the thermal neutron flux distribution change with respect to the amount.

(C) Single-assembly neutron flux discontinuity factor diffusion model Based on the homogeneous diffusion node method using the aggregate neutron flux discontinuity factor, the homogenous neutron flux distribution in the node where each aggregate is homogenized is calculated. Ask. Here, the aggregate discontinuity factor is defined as the ratio of the average neutron flux of each face of the aggregate to the aggregate average neutron flux in the infinite lattice calculation.

In order to calculate the inhomogeneous neutron flux from the homogeneous neutron flux in the node, the “ratio of the inhomogeneous thermal neutron flux to the homogeneous thermal neutron flux at the aggregate boundary” evaluated in the infinite lattice calculation and “in the node” Multiplied by the ratio of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the node is used as a correction factor for the transient component due to thermal neutron diffusion in the homogeneous thermal neutron flux distribution to obtain the inhomogeneous thermal neutron flux in the node. The distribution is calculated, and the nuclear properties of the aggregate are calculated from the inhomogeneous thermal neutron flux distribution. As the homogeneous diffusion model, a multi-group node method, a modified first-group node method, or the like is used.

(D) Single-assembly neutron flux discontinuity factor boundary perturbation model Neutron flux and neutron flow at the node boundary where each aggregate is homogenized based on the homogeneous diffusion node method using the aggregate neutron flux discontinuity factor And ask.

Using the change in the nuclear properties of the aggregate with respect to the change in the neutron flux and the neutron flow at the aggregate boundary evaluated in the infinite lattice calculation as the sensitivity coefficient, Calculate changes in body core properties.

As the homogeneous diffusion model, a multi-group node method,
A modified first group node method, a modified first group difference method, or the like can be used.

[2] In the calculation of the limit power ratio, R
In addition to directly calculating the factor change, it can also be calculated from the fuel rod output change. [3] In the diffusion model, the combustion history effect accompanying the offset bundle adjacency is corrected using the burnup distribution and the spectrum history distribution in the aggregate.

As described above, in the present invention, in the D-lattice core, the core characteristics in consideration of the offset bundle adjacency are calculated based on the table and the diffusion model prepared by the detailed calculation of the multi-assembly, and the combustion history effect is also considered. Therefore, changes in nuclear properties can be calculated with high accuracy. In addition, the critical eigenvalue of the reactor core and the power distribution in the reactor,
The influence on the count value of the neutron detector in the reactor, the irradiation amount of the detector, the irradiation amount of the control rod, etc. can be considered.

FIG. 1 is a block diagram showing the first embodiment of the present invention. Aggregate average nuclear constant calculation means 11
Calculates an aggregate average nuclear constant using a correction amount for an aggregate average nuclear constant or an infinite lattice average nuclear constant that is evaluated in advance by a multi-aggregate detailed calculation according to a combination of an adjacent D lattice bundle and an offset bundle. Things. That is, the node average nuclear constant is calculated as a function of the burnup of the node, the void fraction, and the like, and at this time, the aggregate average kernel constant evaluated in advance by the multi-aggregate detailed calculation according to the combination of the adjacent D lattice bundle and the offset bundle. The numbers are used to take into account the effects of neighboring aggregates.

The in-core power distribution calculation means 12 calculates the core critical eigenvalue, neutron flux distribution, and power distribution based on the core diffusion calculation using the nuclear constants calculated by the aggregate average nuclear constant calculation means 11. Then, the neutron flux distribution in the reactor and the core eigenvalue are calculated based on the diffusion model using the node average nuclear constant. As the diffusion model, for example,
A modified one-group difference model as shown in the document of J. Wooly, a multi-group node method model as shown in the document of K. Smith described later, or the like can be used. The reactor power distribution is calculated based on the neutron flux distribution, and the average power of the cross section of the assembly is calculated.

The local output distribution calculating means 13 calculates
Calculates the fuel rod local output using the correction amount for the fuel rod local output in the fuel assembly or the infinite grid fuel rod local output that has been evaluated in advance by the multi-assembly detailed calculation according to the combination of the lattice bundle and the offset bundle. is there.

Next, the linear power density calculation means 14 uses the assembly cross-sectional average output calculated by the in-core power distribution calculation means 12 and the fuel rod local output calculated by the local power distribution calculation means 13 to generate Calculate the linear power density of.
In the linear power density calculating means 14, the set given by the fuel rod local power distribution evaluated in advance by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and the offset bundle as the thermal limit value of the core The maximum linear power density of each fuel assembly cross-section is calculated using the maximum local power peaking coefficient of the fuel assembly and the average power of the cross-section of the assembly obtained by calculating the power distribution in the core. In this case, the accuracy can be further improved by using the output corrected for the offset bundle adjacent effect as the aggregate cross-sectional average output by the in-core power distribution calculation means 12.

In the limit power ratio calculating means 15, the fuel rods R evaluated in advance by the detailed calculation of the multi-assembly according to the combination of the adjacent D lattice bundle and the offset bundle.
Using the maximum R factor of the assembly obtained from the factor distribution and the average output of the assembly, a critical output ratio, which is a monitoring index of burnout of a fuel rod of each fuel assembly, is calculated. Note that the fuel rod R factor can also be calculated from the fuel rod local output obtained by the local power distribution calculating means 13.

Next, the in-reactor neutron flux meter counting value calculating means 16 calculates the in-reactor neutron meter position evaluated in advance by the multi-assembly detailed calculation in accordance with the combination of the adjacent D lattice bundle and the offset bundle. Using the thermal neutron flux, the in-furnace neutron detector count value and the detector irradiation amount are calculated. Then, the in-furnace power distribution correcting means 17 obtains D
The calculated value of the in-reactor power distribution is corrected by adapting the calculated value of the in-core neutron flux measurement device in which the adjacent effect between the lattice bundle and the offset bundle is corrected to the actually measured value of the in-core neutron flux measurement device. By learning the error of the calculated value of the in-core neutron detector count value corrected in this way with respect to the actually measured value, the in-core power distribution calculated value can also be corrected.

The control rod irradiation amount calculating means 18 uses the control rod position thermal neutron flux evaluated in advance by the multi-aggregate detailed calculation according to the combination of adjacent D lattice bundles and offset bundles. Is calculated.

Next, a multi-aggregate calculation, an average nuclear constant of the aggregate corrected for the offset bundle adjacent effect, an aggregate discontinuity factor, a local output of the fuel rod, a fuel rod R factor, a thermal neutron flux at a neutron measuring instrument position, and the like. The method for calculating the following will be described in detail.

(A) Multi-aggregate detailed two-dimensional calculation model The adjacent effect between the D lattice bundle and the offset bundle is
It is determined by a combination of adjacent D lattice bundles and offset bundles. FIG. 2 shows eight possible patterns for combinations of adjacent bundles centered on the offset bundle. Since the influence of the bundle adjacent to the center bundle of interest from the diagonal diagonal direction is small, it is omitted from the eight patterns here. In FIG. 2, D indicates a D lattice bundle, and S indicates an offset bundle. A multi-aggregate calculation may be performed for each such pattern to evaluate the aggregate core characteristics.

When the combustion proceeds with the assembly displaced,
The burning history effect and the burning degree distribution effect resulting from the change in the neutron spectrum in the assembly due to the change in the water gap width are accumulated, and a difference from the effect when the assembly is displaced instantaneously occurs. In order to incorporate this hysteresis effect, it is necessary to perform a combustion calculation in the multi-aggregate calculation.

FIG. 3 is a characteristic diagram showing, for the multi-aggregate pattern P1 in FIG. 2, the combustion change of the aggregate maximum local peaking coefficient of the D lattice bundle and the offset bundle in comparison with the infinite lattice system. However, it is assumed that the void fraction of each aggregate is equal. As shown in FIG. 3, it can be seen that the amount of change from the calculation of the infinite lattice tends to decrease with combustion.

Although the number of combination patterns of the D lattice and the offset bundle is not so large, the number of combinations actually increases if the differences in the burnup and void ratio of each bundle are taken into account, and the calculation cost increases. However, the combination of the burnups can be approximated by a combination of the burnups for each representative batch type, provided that the offset bundle is introduced stepwise in each cycle.

In order to reduce the error due to the approximation of the combination of the multi-aggregates, instead of giving the aggregate core characteristics itself by the multi-aggregate calculation, only the correction component from the infinite lattice calculation may be given. In the calculation of the core characteristics of the aggregate in each aggregate, first, as a first approximation value, the core characteristics of the aggregate are evaluated by infinite lattice combustion calculation.
Since the correction component of the influence of the adjacent bundle is a second-order effect, the error is not so large even when evaluated by a combination of a typical bundle type, burnup, and void ratio. In this method, only the correction amount from the infinite grid calculation needs to be calculated using a typical combination pattern of multi-aggregates, so that the calculation cost can be reduced.

Further, as a method of simplifying the correction calculation, a multi-aggregate detailed calculation is performed in advance for a combination of the D lattice bundle and the offset bundle in which the thermal limit values such as the maximum linear output density and the limit output ratio are strictest. Using the corrected amount of the fuel rod local power in the fuel assembly and the calculated value of the fuel rod R factor to the infinite lattice calculated value for all D lattice bundles and offset bundles, the core power density and critical power ratio are calculated by core diffusion calculation. Correction calculation can also be performed for each.

(B) Analytical Diffusion Model Next, the change of the thermal neutron flux distribution in the assembly with respect to the variation of the water gap width of the assembly of interest from the infinite lattice system is calculated based on the analytical model for the diffusion equation. Thermal neutron flux change calculation means is provided, and a change in the thermal neutron flux distribution in the aggregate due to a substantial change in the water gap width of the aggregate due to the proximity of the D lattice bundle and the offset bundle is calculated based on an analytical model for the diffusion equation. Ask. And
From this thermal neutron flux distribution, the amount of change in nuclear properties is evaluated.

In this method, for example, for a two-dimensional one-dimensional system in which the water gap and the fuel region inside the channel box are homogenized, the neutron flux distribution when the water gap width is given is analytically calculated. At this time, the analytical expression of the thermal neutron flux change in the fuel region due to the water gap change by the x-direction one-dimensional diffusion model is given by the following expression (1).

[0149]

(Equation 1)

Here, x is the distance from the assembly boundary, κ is the reciprocal of the thermal neutron diffusion distance, and a is the change width of the water gap.

The change in the thermal neutron flux distribution in the assembly can be approximated by the product δΨ (x) δΨ (y) of the one-dimensional distribution in the x and y directions. This method has an advantage that it is not necessary to prepare a table by detailed calculation in advance because it is based on the analysis model.

(C) Single-assembly neutron flux discontinuous factor diffusion model Next, a non-homogeneous thermal neutron flux calculating means for calculating a non-homogeneous thermal neutron flux distribution in the aggregate is provided, and the water of the D lattice bundle and the offset bundle is provided. The effect of different gap widths is considered by using the aggregate neutron flux discontinuity factor obtained by the infinite lattice calculation. Therefore, no multi-aggregate calculations are required. In this method, the neutron flux distribution in the assembly is obtained by the diffusion node method using the neutron flux discontinuity factor for the system consisting of nodes where each assembly is homogenized. To evaluate.

That is, based on the diffusion theory using the neutron flux discontinuity factor, the homogenous thermal neutron flux distribution in the node where each aggregate is homogenized is calculated, and the “heat flux distribution at the aggregate boundary evaluated in the infinite lattice calculation is calculated. The ratio of the ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux multiplied by the ratio of the inhomogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the assembly is multiplied by the transient thermal neutron flux distribution of the homogeneous thermal neutron flux distribution. This is the component correction coefficient. This calculates the heterogeneous thermal neutron flux distribution in the assembly. Then, the aggregate average nuclear constant is corrected using the heterogeneous neutron flux distribution by the heterogeneous thermal neutron flux calculation means to calculate the core eigenvalue, the in-reactor neutron flux distribution, and the power distribution.

Here, the aggregate neutron flux discontinuity factor is defined as the ratio of the average neutron flux on each surface of the aggregate to the aggregate average neutron flux in the infinite lattice system. The multigroup diffusion node method itself using discontinuous factors is already known. For example, see the document "Assembly Homogenization".
Techniques for Light Water Reactor Analysis, "KS
Smith, Progress in Nuclear Energy, vol.17, p303,1986
It is described in.

An example of a method of calculating a local power distribution in an assembly using a neutron flux distribution obtained by the node method calculation is described in “SIMULATE-3 Pin Power Reconstruction M
ethodology and Benchmarking, "KRRempe et.al, Proce
edings of lnternational Reactor Physics Conferenc
e, 111-19, Jackson Hole, 1988.

Since the water gap width differs between the D lattice bundle and the offset bundle, the neutron flux distribution in the aggregate differs, and the neutron discontinuity factor in the aggregate also differs. K.Smi above
As shown in the document of “th”, the neutron flux discontinuity factor gives a boundary condition for the neutron flux at the boundary of each node where the aggregate is homogenized in the diffusion node method.
That is, for example, the following boundary condition for the neutron flux is given at the x-direction boundary between the target aggregate n and the adjacent aggregate m.

[0157]

(Equation 2)

Here, f represents a neutron flux discontinuity factor at the interface, and Ψ represents a neutron flux at the interface. If the target aggregate and the adjacent aggregate are of the same lattice type in the infinite lattice system, the neutron discontinuity factors are also equal, and the above boundary conditions merely represent the continuity of the neutron flux. However, if the neutron discontinuity factor f on the adjacent surface between adjacent aggregates is different, a virtual source of neutrons will be generated at the node boundary based on the above boundary conditions. The change of the neutron flux distribution in the aggregate due to the difference can be calculated. In the multi-group diffusion node method, the whole core calculation using the neutron flux discontinuity factor for the thermal neutron flux,
The node average thermal neutron flux and the node boundary thermal neutron flux are provided.

Next, the neutron flux distribution in the homogenized assembly can be expanded and calculated using the node average neutron flux, the node boundary neutron flux, and the like. As an example, the above K. Rem
In the literature of pe, the two-dimensional distribution of the thermal neutron flux Ψ _{2 in} the homogenized aggregate node is developed in the form of the following equation.

[0160]

(Equation 3)

Here, Ψ ₁ is a fast neutron flux, c _i is an expansion coefficient, and f _i is a sinh and cosh function.

If the ratio of the inhomogeneous neutron flux to the homogeneous neutron flux is known, the inhomogeneous neutron flux can be calculated by multiplying the homogeneous thermal neutron flux distribution of the above equation (3) by this ratio. The diffusion node method assumes that this ratio can be approximated by the ratio of the inhomogeneous neutron flux distribution Ψ ^het and the homogeneous neutron flux distribution で obtained by infinite lattice calculation, as shown in the above-mentioned K. Rempe document. . That is, the inhomogeneous thermal neutron flux distribution Ψ ₂ ^het in the assembly node placed in the core is calculated by the following equation.

[0163]

(Equation 4)

Here, the symbol ∞ indicates an infinite lattice calculated value.

However, when the present inventors applied this method to the offset bundle system, the results of the diffusion node method for the node average neutron flux, the neutron flux at the node boundary and the neutron flow at the node boundary showed that the multi-aggregate detailed calculation Despite good agreement, it was found that the error was large for the fuel rod local output. The reason is that the thermal neutron flux distribution assumes that the ratio of the inhomogeneous neutron flux to the homogeneous neutron flux can be approximated by the ratio of the inhomogeneous neutron flux distribution and the homogeneous neutron flux distribution obtained by the infinite lattice calculation. Was not satisfied.

In order to accurately calculate the inhomogeneous thermal neutron flux distribution from the in-node homogeneous thermal neutron flux, the present inventors evaluated the “homogeneous thermal neutron flux of the inhomogeneous thermal neutron flux at the aggregate boundary evaluated in the infinite lattice system. By multiplying the ratio of the neutron flux to the neutron flux by the ratio of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the aggregate, the factor is used as the correction factor for the transient component of the thermal neutron flux distribution due to thermal neutron diffusion. It was found that the distribution of inhomogeneous thermal neutron flux in the assembly node can be calculated with high accuracy.

The transient component of the homogeneous thermal neutron flux distribution due to thermal neutron diffusion is given by the second term of equation (3). Therefore, in this method, the heterogeneous thermal neutron flux is given by the following equation (5) instead of the equation (4).

[0168]

(Equation 5)

Here, the coefficient W representing the “ratio of the inhomogeneous thermal neutron flux to the homogeneous thermal neutron flux at the aggregate boundary” can be expressed by the following equation by the weighted average of the four aggregate boundary surfaces.

[0170]

(Equation 6)

When the heterogeneous thermal neutron flux distribution in the assembly is calculated, the aggregate average nuclear constant, which is a function of the heterogeneous neutron flux distribution, is calculated, and the in-reactor power distribution corrected for the offset bundle adjacency is calculated. . In addition, the fuel rod local power distribution is immediately calculated from the heterogeneous thermal neutron flux distribution in the assembly,
From this, the linear power density of the aggregate can be calculated.

Further, the fuel rod R-factor distribution can be calculated from the fuel rod local power distribution, from which the critical power ratio can be calculated.

The aggregate neutron flux discontinuity factor on each surface of the aggregate is prepared for each aggregate in advance by a single aggregate detailed calculation (infinite lattice calculation using reflection boundary conditions). FIG. 4 shows an example of a function form with respect to the offset amount of the channel box of the aggregate discontinuity factor for the thermal neutron flux.

As an example of the calculation accuracy of this method, FIG.
As with, the control rod is placed around the D lattice bundle
FIG. 5 shows the result of calculating the local output distribution in the aggregate when the offset bundles of the body are adjacent to each other by the present method and comparing the distribution with the reference solution by the multi-aggregate detailed calculation. In FIG. 5, the upper part shows the detailed calculation, the middle part shows the method according to the present invention, and the lower part shows the difference. According to the reference calculation, the aggregate maximum local peaking factor is about 10 due to the neighbors of the offset bundle.
%, But the method reproduces this well.

In this method, the detailed calculation of the multi-assembly bundle is not required when preparing the table of the neutron flux discontinuity factor of the assembly, and the neutron flux discontinuity factor of each surface of the fuel assembly is determined by each surface of the assembly of interest. May be calculated independently as a function of As described above, the present method has an advantage that the preparation and reference of the table are simpler than the above-described method requiring the multi-aggregate calculation.

In the method based on the modified first group diffusion model as the diffusion model, only the fast neutron flux distribution is solved in the whole core calculation. The thermal neutron flux is calculated using the fast neutron flux obtained by the core calculation and the "ratio of thermal neutron flux to fast neutron flux (spectral index)" in an infinite lattice system. If the spectral indices differ between adjacent aggregates, spatial movement of thermal neutrons will occur, and the change of the thermal neutron flux from the infinite lattice will result in a system consisting of nodes where the focused aggregate and the adjacent aggregate are homogenized. On the other hand, it is calculated by applying a diffusion model.

In the modified first group node method, a homogeneous thermal neutron flux in the node is obtained by analytically solving the diffusion equation of this system. Details of this method are described, for example, in the document "Verifica
tionof LOGOS Nodal Method with Heterogeneous Burnu
p Calculations for a BWRcore, "TI wamoto et al., Tr
ansaction of American Nuclear Society, vol. 71, p251,
1994.

Also in the modified one-group method, the change in the thermal neutron flux due to the channel box deformation is caused by the neutron flux discontinuity factor in the boundary condition for the thermal neutron flux in a system consisting of nodes in which the target aggregate and the adjacent aggregate are homogenized. Is used, the calculation can be performed according to the same principle as in the above-described multi-group node method.

(D) Single-assembly neutron flux discontinuity factor boundary perturbation model Another method using the neutron flux discontinuity factor is a combination with the boundary perturbation method. This is based on the diffusion theory calculation using neutron flux discontinuity factor.The node boundary value calculation means for calculating the neutron flux and neutron flux at the node boundary surface where each aggregate is homogenized is provided. The neutron flux and the neutron flow at the boundary of the aggregate are obtained by using the difference of the aggregate discontinuity factor. Calculate changes in properties. This is because, as described above, the diffusion node method using the neutron flux discontinuity factor focuses on the fact that the neutron flux and the neutron flow at the aggregate boundary are accurate.

An example of calculating local peaking in an assembly using “neutron flow / neutron flux” as a perturbation amount at the assembly boundary based on the boundary perturbation method is described in the document “A Boundary Con
dition Perturbation Method for Prediction of Pin P
ower Distribution in LightWater Reactor, "F. Rahnema
et al., Proceedings of Topical Meeting on Reactor
Physics and Shielding, Chicago, 1984.

In this method, for example, the local output LPF (x, y) in the assembly for the fuel rod (x, y) is calculated by the following equation (8).

[0182]

(Equation 7)

The sensitivity coefficient F is prepared by a detailed calculation of a single aggregate or a detailed calculation of a multi-aggregate in which boundary conditions are changed in advance.

However, in the above-mentioned Rahnema reference,
When calculating the neutron flux / flux at the interface based on the diffusion theory, the local neutron near the water gap as in the case of the adjacent offset bundle is not used because the aggregate neutron flux discontinuity factor is not used. It is not possible to calculate changes in neutron flow / flux due to flux distribution changes.

Therefore, the node boundary value r is multi-grouped by using a neutron flux discontinuity factor for a system consisting of nodes in which a set of interest and an adjacent set are homogenized, as in the aforementioned diffusion node method. Determined by applying the diffusion node method. Alternatively, in the modified one-group difference method, the node boundary thermal neutron flux may be calculated from the spectral index of the adjacent node and the neutron flux discontinuity factor using an empirical load factor. In this case, the thermal neutron flux Ψ _{2n at} the boundary between the node m and the node n is expressed by the following equation (9).

[0186]

(Equation 8)

Equation (8) shows the correction calculation method using the boundary perturbation method for the local output of the aggregate. The correction calculation of the nuclear characteristics of other aggregates is also performed by the neutron flux / neutron flux of each nuclear characteristic. Can be calculated in a similar manner using the sensitivity coefficient for

If the combustion proceeds while the fuel assembly is displaced, the hysteresis effect of the neutron spectrum change in the assembly and the effect of the burnup distribution due to the water gap width change are accumulated, and the assembly is instantaneously displaced. A difference from the effect of
This is because the neutron spectrum (the ratio of the thermal neutron flux to the fast neutron flux) changes from an infinite system, causing a delay in uranium combustion and an increase in plutonium accumulation, and is called a spectrum hysteresis effect.

In the diffusion node method, the combustion history effect cannot be directly considered, as in the combustion calculation by the multi-aggregate detailed calculation. In the core performance calculation by the diffusion node method,
As a method of correcting the spectrum hysteresis effect due to the spectrum interference when fuel assemblies having different neutron spectra are adjacent to each other, for example, a method described in Japanese Patent Application No. 6-210243 “Method and Apparatus for Reactor Core Performance Calculation” As described above, there is a method of correcting a spectrum history, which is an average burnup of a ratio of a neutron spectrum in a reactor core to a spectrum in an infinite lattice system, as a parameter.

In the present invention, this technique is applied to the spectrum hysteresis effect due to the offset bundle adjacency, and the burnup distribution and the spectrum hysteresis distribution in the fuel assembly in which the change effect of the thermal neutron flux distribution due to the change of the water gap width is corrected are calculated. Then, the correction amount of the average nuclear constant of the aggregate, the fuel rod local power distribution, and the like due to the combustion history effect is calculated. For example, regarding the fuel rod local output, the fuel rod position (x,
The spectrum history SH (x, y) in y) is expressed by the following equation (10).

[0191]

(Equation 9)

Here, E is the burnup, Ψ ₂ and Ψ ₁ are the thermal neutron flux and the fast neutron flux at the position (x, y) calculated in consideration of the offset bundle adjacency, respectively, and ∞ is infinite. Indicates the value in the lattice system. The correction of the local power distribution LPF in consideration of the spectrum history is based on the thermal group fission cross section Σ
By using the spectrum history correction coefficient (∂Σ / ∂SH) for the following equation (11).

[0193]

(Equation 10)

Similarly, by offset bundle adjacency,
The hysteresis effect caused by the burnup distribution in the node can also be corrected using the sensitivity coefficient related to the burnup as described in the above-mentioned document.

Next, a second embodiment for the deformation of the channel box will be described. Accurate calculation of core core characteristics in consideration of channel box deformation is achieved by the following processes [1] to [5].

[1] The fast neutron irradiation amount or burnup on each surface of the channel box is integrated by core performance calculation, and the irradiation growth amount on the opposite surface of the channel box is a function of the fast neutron irradiation amount or burnup on each surface. And the amount of deformation of the channel box is calculated from the difference in the amount of irradiation growth on the opposing surface. [2] Considering the channel box deformation between the focused assembly and the adjacent aggregate, aggregate nuclear characteristics such as average nuclear constant of the aggregate, local power of the fuel rod, thermal neutron flux at the neutron detector position, thermal neutron flux at the control rod position, etc. Is calculated. [3] The calculation of the above-mentioned change in the nuclear properties of the aggregate from the deformation amount of the channel box is based on the following detailed calculation model of the multi-aggregate or the neutron diffusion model.

(A) Multi-assembly detailed two-dimensional calculation model The multi-assembly detailed two-dimensional calculation for the target assembly and the adjacent assembly allows the average nuclear constant of the assembly, fuel rod output distribution, fuel rod R-factor distribution, etc. The change amount of the core property of the aggregate is directly prepared as a table of the deformation amount of the water gap width of each surface of the aggregate of interest. Here, the multi-assembly detailed two-dimensional calculation is a detailed calculation that takes into account the inhomogeneity of each assembly as it is for a system that simulates a part of the core by combining fractions (usually 4 to 16). Is performed, the effect of the adjacent aggregate on the aggregate of interest is evaluated in detail.

(B) Single-assembly neutron flux discontinuity factor diffusion model The amount of change in the aggregate neutron flux discontinuity factor is calculated from the amount of channel box deformation. The thermal neutron flux distribution in the assembly is obtained based on the diffusion model, and the amount of change in nuclear properties is evaluated from the thermal neutron flux distribution. Here, the aggregate discontinuity factor is defined as the ratio of the average neutron flux of each face of the aggregate to the aggregate average neutron flux in the infinite lattice system. The amount of change in the neutron flux discontinuity factor of the aggregate is prepared in advance as a table of the amount of channel box deformation by a single aggregate detailed calculation.
The homogeneous diffusion model includes a multi-group node method, a modified 1
A group node method or the like can be used.

(C) Single-assembly neutron flux discontinuity factor boundary perturbation model Neutron flux and neutron flow at the node boundary where each aggregate is homogenized based on the homogeneous diffusion node method using the aggregate neutron flux discontinuity factor Ask for.

Using the neutron flux and the neutron flow at the node boundary as described above, as a sensitivity coefficient, the change in the neutron flux and the neutron flow at the aggregate boundary evaluated in the infinite lattice system is used as a sensitivity coefficient. Calculate changes in nuclear properties.

As the homogeneous diffusion model, a multi-group node method,
A modified first group node method, a modified first group difference method, or the like can be used.

(D) Analytical diffusion model The change of the thermal neutron flux distribution in the assembly is obtained from the channel box deformation based on the analytical solution of the two-region diffusion model in which the water gap and the fuel region are homogenized. The amount of change in nuclear properties is evaluated from the flux distribution.

[4] In the calculation of the limit power ratio in consideration of the channel box deformation, in addition to directly finding the change in the R factor in the assembly, the calculation can also be made from the change in the fuel rod output. [5] Combustion history effect due to channel box deformation
The correction is made using the burn-up distribution in the aggregate due to the channel box deformation and the spectrum history distribution defined as the burn-up average value of the ratio of the thermal neutron flux to the fast neutron flux.

As described above, in the second embodiment,
Since the fast neutron irradiation or burnup on each side of the channel box is integrated by core performance calculation, it is possible to predict and evaluate the amount of channel box deformation due to neutron irradiation and calculate the core characteristics taking into account the channel box deformation Becomes

Further, the calculation of the core characteristics in consideration of the channel box deformation is performed based on the table and the diffusion model prepared by the multi-aggregate detailed calculation from the channel box deformation amount of the target aggregate and the adjacent aggregate. Since the hysteresis effect of channel box deformation is also taken into account, it is possible to accurately calculate changes in nuclear characteristics.

In addition, the influence of the channel box deformation on the critical eigenvalue of the reactor core, the power distribution in the reactor, the count value of the in-core neutron detector and the detector radiation, the control rod radiation, etc., which have been neglected in the past, is also taken into consideration. it can.

FIG. 6 is a block diagram showing a second embodiment of the present invention. The in-core power distribution calculating means 19 receives the state data of the core and calculates the neutron flux distribution in the reactor and the core eigenvalue based on the diffusion model. As the diffusion model, for example, a modified one-group difference model as shown in the above-mentioned J. Wooly document, a multi-group node method model as shown in the K. Smith document, and the like are used. The reactor power distribution is calculated based on the neutron flux distribution, and the average power of the cross section of the assembly is calculated.

The channel box dose calculating means 20 calculates the fast neutron dose on each surface of the channel box of the assembly based on the neutron flux distribution. Here, the fast neutron flux Ψ _n on the surface n is explicitly calculated by the multi-group diffusion node method. In the modified one-group difference model, the node average neutron flux of the focused set and the surrounding set is calculated using a polynomial. It can be calculated by interpolation. Thus, the fast neutron irradiation amount φ can be calculated as shown in Expression (12) by time-integrating the fast neutron flux. Where d
t is a time width.

[0209]

[Equation 11]

The channel box deformation amount calculating means 21 evaluates the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount on each surface. As an evaluation method, it can be calculated from the irradiation growth amount in the axial direction on the opposing surface of the channel box (claim 25).

In the channel box deformation amount calculation method,
The channel box deformation is analytically calculated assuming that the axial bending of the channel box has a uniform curvature according to the difference between the irradiation growth amounts on the two opposing surfaces A and B (claim 33). That is, as shown in FIG. 7, the average fast neutron doses in the axial direction of the surfaces A and B of the channel box are φ _A and φ _B respectively, and the curvature of the channel box in the axis (Z) direction is the curvature. Is small, and is expressed by the following equation (13). Then, the bending amount at the axial height Z is expressed by Expression (14) by integrating Expression (13). Therefore, the amount of bending when the upper and lower ends are fixed is expressed by equation (15). Further, the irradiation growth function is expressed as in the equations (16) and (17). FIG. 8 shows an example of this function form.

[0212]

(Equation 12)

Here, a, b, c, d, and e are fitting coefficients. This coefficient can be obtained from an empirical formula. In addition, this coefficient is determined using the past measured value of the channel box deformation amount in order to correct random bending due to release of residual stress at the time of channel manufacturing, error due to variation in irradiation growth of Zircaloy, etc. (Claim 42).

As a method of calculating the amount of channel box deformation, there is also a method based on the finite element method in addition to the analytical method of the above equation (15). As shown in FIG. 9, the channel is divided into axial nodes, and the lengths lA ⁱ and lB ⁱ after irradiation growth of both A and B facing the channel are obtained for each node i. After the nodes are stacked in the axial direction, the distance between each node and a straight line connecting the upper and lower ends of the A surface or the B surface of the channel can be obtained as a bending amount δ ⁱ .

The fast neutrons that cause irradiation growth are neutrons having an energy of 1 MeV or more, and do not always match the total fast neutrons given in the core performance calculation. Therefore, the ratio f of the fast neutrons used in the calculation of the fast neutron dose to the total fast neutrons is expressed by the following equation (18).
It can be calculated as a table of the burnup E and the void fraction V (claim 31).

[0216]

(Equation 13)

Further, in the second embodiment, the core performance calculation device is configured by using an irradiation growth model using the one-to-one correspondence between the fast neutron irradiation amount and the burn-up (claims 26 to 27). Item 30). In this case, a burn-up distribution calculating means is provided in place of the channel box dose calculating means 20 of FIG.

Next, a method of calculating the burnup on each surface of the channel box of the aggregate by the burnup distribution calculation means will be described. Here, the burnup En on the surface n
Is calculated by time-integrating the output density at the channel box position since the output distribution in the node is explicitly calculated in the multi-group diffusion node method. Further, in the modified one-group difference model, it can be calculated by extrapolating the node average burnup of the target aggregate and the peripheral aggregate using a polynomial.

In the model using the burnup (Claim 31), the irradiation growth distortion amount of the channel box due to neutron irradiation is calculated as a function of the burnup E of each surface by, for example, the following (1)
It can be calculated as in equations 9) and (20).

[0220]

[Equation 14]

Here, a, b, c, d, and e are fitting coefficients. This coefficient can be obtained from an empirical formula. Further, this coefficient may be determined using an actual measured value of the channel box deformation amount in the past in order to correct an error depending on uncertainty or the like at the time of manufacturing the channel (claim 42). FIG. 10 shows an example of the function form of ε (E).

Next, the in-furnace power distribution recalculating means 22
Channel box deformation amount calculation means 2 as described above
Using the amount of channel box deformation evaluated in step 1,
Calculate the assembly nuclear constant and neutron flux discontinuity factor corrected for the channel box deformation effect, and use them to recalculate the critical eigenvalue of the core, the reactor power distribution, and the fuel assembly cross-sectional average power distribution. 20, claim 26).

The cotton power density calculation means 23 calculates the maximum local power peaking coefficient of the aggregate given by the fuel rod local power distribution corrected for the channel box deformation effect and the set obtained by calculating the power distribution in the core as the thermal limit value of the core. The maximum linear power density of each fuel assembly section is calculated using the body section average output (claims 21 and 27). In addition, instead of obtaining the maximum local output peaking coefficient of the aggregate from the fuel rod local output distribution corrected for the channel box deformation effect, the accuracy is reduced, but the channel is directly calculated with respect to the maximum local output peaking coefficient of the aggregate. The box deformation effect may be corrected.

Further, the limit power ratio calculating means 24 uses the aggregate maximum R factor obtained from the fuel rod R factor distribution in consideration of the channel box deformation and the average power of the aggregate to calculate the fuel of each fuel assembly. A limit power ratio, which is an index for monitoring burnout of the rod, is calculated (claims 22 and 28). In addition,
The fuel rod R-factor distribution can also be calculated from the fuel rod local power distribution described above. Also, instead of obtaining the maximum R factor of the aggregate from the fuel rod R factor distribution corrected for the channel box deformation effect, the accuracy is reduced, but the channel box deformation effect is directly applied to the maximum R factor of the aggregate. It may be corrected.

In the calculation of the linear power density, the critical power ratio, etc., the average output of the fuel assembly used in the calculation is
Accuracy can be improved by using the average output of the fuel assembly recalculated using the nuclear constant in which the channel box deformation is corrected (claims 37 and 38).

Next, the in-reactor neutron flux detector counting value calculating means 25 uses the thermal neutron flux at the in-reactor neutron measuring device position in consideration of the channel box deformation to calculate the in-reactor neutron detector count value, the irradiation amount, and the like. Is calculated (claims 23 and 29). An example of a method of calculating the count in the case of a movable in-core neutron flux detector (TIP) as the in-core neutron flux detector is described in the aforementioned Wooly document. The detector irradiation amount is given by the time integration of the thermal neutron flux at the position of the neutron detector in the reactor.

Although not shown in FIG. 6, the error of the calculated value of the in-reactor neutron detector count corrected for the channel deformation with respect to the actually measured value is learned based on the method of the above-mentioned Tuiki literature, etc. The in-furnace power distribution calculation value can be corrected (claim 36). Further, the control rod irradiation amount calculating means 26 calculates the control rod irradiation amount by time integration of the control rod position thermal neutron flux in which the channel box deformation effect is corrected (claims 24 and 30).

Next, three different methods for calculating the aggregate average nuclear constant, the aggregate discontinuity factor, the fuel rod local output, the fuel rod R factor, the thermal neutron flux at the neutron measuring instrument position, etc., in which the channel box deformation effect is corrected. The method will be described in detail.

(A) Detailed two-dimensional calculation model of multi-aggregate First, as method 1, aggregate average nuclear constant, fuel rod local output, fuel rod R-factor, neutron for channel box deformation of target aggregate and adjacent aggregate Nuclear property changes such as thermal neutron flux at the measurement device position are prepared as a table or a fitting equation, and the nuclear property changes in the fuel assembly in which the channel box deformation effect is corrected are directly calculated. The table or the fitting formula of the change of the characteristic amount of the focused aggregate with respect to the channel box deformation amount of the focused aggregate and the adjacent aggregate is evaluated by multi-aggregate two-dimensional detailed calculation (claim 32).

Here, the multi-assembly detailed two-dimensional calculation refers to a system in which a part of the core is simulated by combining fractions (usually 4 to 16) with the inhomogeneity of each assembly taken into account. This is a method for evaluating in detail the influence of the adjacent aggregate on the aggregate of interest by performing the detailed calculation described above.

This table or fitting formula is a function of not only the amount of deformation of each surface of the aggregate of interest, but also the amount of deformation of the adjacent surface of an adjacent aggregate, and is generally a complex function for a combination of multivariates. On the other hand, Japanese Patent Application Laid-Open No. H4-110698 describes that the calculation is simply made from the displacement amount of the fuel assembly, and the evaluation method is not clear at all. In this method, this function is arranged as a table of the water gap width for each surface of the channel box of the aggregate of interest. This table is also a multivariate function (four-variable function) for the water gap width of each surface,
For example, the following method can be considered as a relatively simple reference method.

First, as a first step, it is assumed that there is no channel box deformation of an aggregate adjacent to the target aggregate, and a case is considered where only the channel box 2b of the target aggregate is deformed. As shown in FIG. 11, the water gap deformation amount δxl (i) on the left side in the x direction and the water gap deformation amount δxr (i) on the right side in the x direction of the channel box 2b of the target aggregate are equal (the absolute values are equal). (Opposite in sign) δx. Similarly, it is assumed that the amount of water gap deformation δyt (i) on the upper side in the y direction of the focused aggregate channel box 2b is equal to the amount of water gap deformation δyb on the lower side in the y direction. The change in the nuclear properties of the aggregate due to the deformation of the channel box is represented by v (δx,
δy) is arranged and referred to as a two-dimensional table or a fitting equation.

Next, as a second step, correction is made in consideration of the influence of the change in the water gap width due to the deformation of the adjacent surface of the aggregate channel box adjacent to the aggregate of interest. FIG.
In FIG. 1, the position of the channel box after the deformation is indicated by a dotted line. In this case, for example, the water gap deformation amount δxl on the left side in the x direction of the focused aggregate channel box 2b
Is the deformation amount δxr of the left surface of the focused aggregate channel box 2b and the right surface of the left adjacent aggregate channel box 2a.
(I-1). The same applies to the water gap deformation amount δxr on the right side in the x direction.

Since the amount of deformation of the aggregate channel boxes on both sides is generally different, δxl (i) ≠ δxr (i) when the deformation of the aggregate channel boxes on both sides is taken into account.
Becomes Here, it is assumed that dx = δxl−δxr. Similarly, in the y direction, dy = δyt−δyb. Therefore, the table is first referred to as v (δxl, δyt),
The correction amount for this is arranged as w (dx, dy),
See also Finally, the nuclear property change is given by v + w.

FIG. 12 shows an example of the function form of the aggregate maximum local peaking coefficient with respect to δx when δx = δy and δy = 0. FIG. 13 shows an example of a function form with respect to dx of the correction amount of the aggregate maximum local peaking coefficient when dx = dy and dy = 0.

(B) Single-assembly neutron flux discontinuous factor diffusion model Next, method 2 will be described. In this method 2, the amount of change in the neutron flux discontinuity factor of the aggregate is calculated from the amount of channel box deformation of the aggregate of interest and the adjacent aggregate, and the neutron flux discontinuity factor is calculated for a system in which each aggregate is homogenized. The neutron flux distribution in the aggregate is obtained by the diffusion node method used, and the amount of change in the nuclear characteristics of the aggregate is evaluated from the neutron flux distribution (claim 33). Here, the aggregate neutron flux discontinuity factor is defined as the ratio of the average neutron flux of each face of the aggregate to the aggregate average neutron flux in the infinite lattice system.

As described above, the multi-group diffusion node method calculation method using discontinuous factors is described in the document "Assembly Homogeni".
zation Techniques for Light Water Reactor Analysi
s, "KSSmith, Progress in Nuclear Energy, vol.17, p30
3,1986. An example of a method of calculating a local power distribution in an assembly using a neutron flux distribution obtained by a node method calculation is “SIMULATE-3 Pin Power Rec.
onstruction Methodology and Benchmarking, "KRRemp
e et.al, Proceedings of International Reactor Physi
cs Conference, III-19, Jackson Hole, 1988.

When the channel box is deformed, the neutron flux distribution in the aggregate changes due to the change in the water gap width.
Aggregate neutron discontinuity factors also change. As shown in the above-mentioned K. Smith document, the neutron flux discontinuity factor gives a boundary condition for the neutron flux at the boundary of each node where the aggregate is homogenized in the diffusion node method. That is, for example, at the boundary in the x direction between the target assembly n and the adjacent assembly m, the boundary condition of Expression (2) for the neutron flux is given. Here, in equation (2), f represents a neutron flux discontinuity factor at the interface, and Ψ represents a neutron flux at the interface.

If the target aggregate and the adjacent aggregate have the same lattice type in the infinite lattice system, the neutron discontinuity factors are also equal, and the above boundary conditions merely represent the continuity of the neutron flux. However, if the neutron discontinuity factor f at the adjacent surface between the adjacent aggregates differs due to the channel box deformation, a virtual source of neutrons will be generated at the node boundary based on the above boundary conditions, and thus the difference in the aggregate discontinuity factor will occur. From, the change of the neutron flux distribution in the aggregate due to the channel box deformation can be calculated. In the multigroup diffusion node method, a node average thermal neutron flux and a node boundary thermal neutron flux are provided by a total core calculation for a thermal neutron flux using a neutron flux discontinuity factor.

Next, the neutron flux distribution in the homogenized aggregate can be expanded and calculated using the node average neutron flux, the node boundary neutron flux, and the like. As an example, the above K.Remp
In the document of e, the two-dimensional distribution of the thermal neutron flux Ψ _{2 in} the homogenized aggregate node is expanded in the form of the equation (3). Here, in equation (3), Ψ ₁ is a fast neutron flux,
c _i is an expansion coefficient, and f _i is a sinh, cosh function.

Once the thermal neutron flux distribution in the homogenized assembly is obtained, the non-homogeneous thermal neutron flux distribution in the assembly is synthesized by combining this with the non-homogeneous neutron flux distribution obtained by the infinite lattice calculation. Can be calculated. The aggregate average nuclear constant, which is a function of the heterogeneous neutron flux distribution, and the fuel rod local power distribution can be calculated immediately from this. The fuel rod R factor distribution can be calculated from the fuel rod local power distribution.

The amount of change of the neutron flux discontinuity factor with respect to the amount of channel box deformation on each surface of the aggregate is:
It is prepared in advance as a table of channel box deformation amount by a single assembly detailed calculation (infinite grid calculation). Water gap width change δx of aggregate discontinuity factor for thermal neutron flux
FIG. 14 shows an example of a function form with respect to.

As an example showing the calculation accuracy of this method, FIG. 2 showing the effect of the channel box deformation on the local output distribution.
As in 1, the channel box of the aggregate of interest is x, y
In the case where the deformation is performed by 2 mm in the direction and there is no deformation in the adjacent aggregate, that is, when δx = δy = 2 mm, the local output distribution in the aggregate is calculated by this method, and the reference solution by the multi-aggregate detailed calculation and The comparison result is the same as the case shown in FIG. The fuel rod local power calculation method is described in K.
The same method as that described in the above-mentioned document was used, but the above-mentioned non-homogeneity correction of the aggregate was added to the change in the thermal neutron flux distribution. According to the reference calculation, the maximum local peaking coefficient of the aggregate changes by about 10% due to the channel box deformation, but the present method reproduces this well.

According to this method, the detailed calculation of the multi-aggregate is not required when preparing the table of the amount of change of the neutron flux discontinuity factor from the amount of channel box deformation. When calculating the change in the nuclear properties of the assembly from the box deformation, the neutron flux discontinuity factor of each surface of the fuel assembly can be calculated independently as a function of only the deformation of each surface of the channel box of the assembly of interest. Good. As described above, the method 2 has an advantage over the method 1 in that the table can be easily prepared and referred to regardless of the combination of the displacement amounts of the respective surfaces.

In the method based on the modified first group diffusion model (claim 40), only the fast neutron flux distribution is solved in the whole core calculation. The thermal neutron flux is calculated from the fast neutron flux obtained by core calculation and the ratio of thermal neutron flux to fast neutron flux (spectral index) in an infinite lattice system.
Calculate using If the spectral indices differ between adjacent aggregates, spatial movement of thermal neutrons will occur, and the change of the thermal neutron flux from the infinite lattice will result in a system consisting of nodes where the focused aggregate and the adjacent aggregate are homogenized. On the other hand, it is calculated by applying a diffusion model. In the modified first group node method, a homogeneous thermal neutron flux in the node is obtained by analytically solving the diffusion equation of this system. Details of this method can be found in the document "Verification of LOGOS Nodal"
Methodwith Heterogeneous Burnup Calculations for
a BWR core, "T.lwamoto et al., Transaction of Americ
an Nuclear Society, vol. 71, p251, 1994.

Also in the modified one-group method, the thermal neutron flux change due to the channel box deformation is caused by the neutron flux discontinuity factor in the boundary condition for the thermal neutron flux in a system consisting of nodes in which the target aggregate and the adjacent aggregate are homogenized. Is used, the calculation can be performed according to the same principle as in the above-described multi-group node method. Instead of solving the diffusion equation, the thermal neutron flux and the thermal neutron flow can also be calculated from the spectral index of an adjacent node using an empirical load factor, as is well known.

(C) Single-assembly neutron flux discontinuity factor boundary perturbation model As another method using the neutron flux discontinuity factor, a method combined with the boundary perturbation method is used (claim 34). In this method, the neutron flux and neutron flow at the boundary of the aggregate are obtained by using the difference in the discontinuity factor of the aggregate due to the difference in water gap width. Using the sensitivity coefficient, the change in the nuclear properties of the aggregate is calculated. This is because, as described above, the diffusion node method using the neutron flux discontinuity factor focuses on the fact that the neutron flux and the neutron flow at the aggregate boundary are accurate.

An example of calculating local peaking in an assembly by using “neutron flow / neutron flux” as a perturbation amount at the assembly boundary based on the boundary perturbation method is described in “A Boundary Cond
ition Ferturbation Method for Prediction of Pin Po
wer Distribution in LightWater Reactors, "F. Rahnema
et a1., Proceedings of Topical Meeting on Reactor
Physics and Shielding, Chicago, 1984.

In this method, for example, the fuel rod (x, y)
The local output LPF (x, y) in the set with respect to is calculated by the above-mentioned equation (8). The sensitivity coefficient F in the equation (8) is prepared by a single-aggregate detailed calculation or a multi-aggregate detailed calculation in which boundary conditions are changed in advance.

However, in the above-mentioned Rahnema reference,
When calculating the neutron flux / flux at the interface based on the diffusion theory, the local neutron near the water gap as in the case of the adjacent offset bundle is not used because the aggregate neutron flux discontinuity factor is not used. It is not possible to calculate changes in neutron flow / flux due to flux distribution changes.

In the present invention, the node boundary value r is multiplied by using a neutron flux discontinuity factor for a system composed of nodes each having a target aggregate and an adjacent aggregate homogenized, as in the aforementioned diffusion node method. It can be obtained by applying the group diffusion node method (claim 34). Alternatively, in the modified one-group difference method, the node boundary thermal neutron flux may be calculated from the spectral index of the adjacent node and the neutron flux discontinuity factor using an empirical load factor (claim 40). In this case, the thermal neutron flux Ψ _{2n at} the boundary between the node m and the node n is represented by the above-mentioned equation (9).

The above equation (8) shows the correction calculation method by the boundary perturbation method for the local output of the aggregate according to claim 27, and the other kernels according to claims 26 to 30 are described. The correction calculation of the characteristic amount can also be performed by a similar method using the sensitivity coefficient for the neutron flow / neutron flux of each nuclear characteristic amount.

(D) Analytical diffusion model Next, method 3 will be described. This method 3 is different from method 2
Instead of using the diffusion node method to find the thermal neutron flux in the aggregate, the change in the thermal neutron flux distribution in the aggregate based on the analytical model for the diffusion equation is calculated based on the channel box deformation of the aggregate of interest and the adjacent aggregate. Then, the amount of change in nuclear properties is evaluated from the thermal neutron flux distribution (claim 35). In this method 3, for example, for a two-dimensional one-dimensional system in which the water gap and the fuel region inside the channel box are respectively homogenized, the neutron flux distribution when the water gap width is given is analytically calculated.

At this time, the analytical expression of the change in the thermal neutron flux in the fuel region due to the change in the water gap by the one-dimensional diffusion model in the x direction is given by the above-mentioned expression (1). In the equation (1), x is the distance from the boundary of the aggregate, κ is the reciprocal of the thermal neutral prediffusion distance, and a is the water gap change width. The change in the thermal neutron flux distribution in the assembly can be approximated by the product δΨ (x) δΨ (y) of the one-dimensional distribution in the x and y directions. This method 3
Has the advantage that it is not necessary to prepare a table in advance because it is based on an analysis model.

If the combustion proceeds with the assembly displaced,
The hysteresis effect of the neutron spectrum change and the burn-up distribution hysteresis effect in the aggregate due to the water gap width change are accumulated, and a difference from the effect when the aggregate is displaced instantaneously occurs.
A phenomenon in which the neutron spectrum (ratio of thermal neutron flux to fast neutron flux) changes from an infinite system, resulting in a delay in uranium combustion and an increase in the accumulation of plutonium is called a spectral hysteresis effect. In addition, the effect that the uranium combustion delay occurs due to the burnup distribution in the aggregate is called a one-sided burning effect. Since these effects generally work in a direction to offset the instantaneous effects, it is necessary to appropriately consider these effects so as not to take an excessive margin in calculating the thermal limit value.

The amount of channel box deformation increases with the burning of the aggregate, and the spectrum change due to the channel deformation also changes with the burning. For this reason, when the combustion history effect of the channel deformation is evaluated by the detailed combustion calculation using the multi-assembly, the deformation amount may be changed for each combustion step, but the calculation complexity increases.

In order to avoid this, there is a method in which only the instantaneous effect is evaluated in the multi-assembly calculation, and the spectrum hysteresis effect and the one-sided burning effect are corrected by the core performance calculation. In the case of the diffusion node method using discontinuous factors, the channel deformation effect can be calculated only as an instantaneous effect.
Correct by core performance calculation.

In the core performance calculation, as a method of correcting the spectrum hysteresis effect due to the spectrum interference when the fuel assemblies having different neutron spectra are adjacent to each other, for example, Japanese Patent Application No. 6-210243, “Reactor Core Performance Calculation As described in “Method and Apparatus,” there is a method of correcting a spectrum history, which is an average burnup value of a ratio between a neutron spectrum in a reactor core and a spectrum in an infinite lattice system, as a parameter.

Therefore, in the present invention, this method is applied to the spectrum history effect by the channel box deformation, and the burnup distribution and the spectrum history distribution in the fuel assembly in which the channel box deformation effect is corrected are calculated.
A correction amount such as an aggregate average nuclear constant and a fuel rod local power distribution due to the combustion history effect of the channel deformation is calculated (claim 39). For example, regarding the local output of the fuel rod, the spectrum history SH at the fuel rod position (x, y) in the assembly is obtained.
(X, y) is represented by the above equation (10). In the equation (10), E is the burnup, Ψ ₂ and Ψ ₁ are the thermal neutron flux and the fast neutron flux at the position (x, y) calculated in consideration of the channel box deformation, respectively, and ∞ is infinite. Indicates the value in the lattice system. The correction of the local power distribution LPF in consideration of the spectrum history is performed by using a spectrum history correction coefficient (∂Σ / ∂SH) for the thermal group fission cross section Σ.
It is given by the above equation (11).

Similarly, a hysteresis effect caused by a burnup distribution in a node due to a change in thermal neutron flux distribution due to channel deformation can also be corrected using a sensitivity coefficient relating to burnup, as described in the above-mentioned document. it can.

[0261]

As described above, according to the core performance calculating apparatus of the present invention, the calculation of the core characteristics taking into account the effect of the adjacent D lattice bundle and offset bundle is performed by the D lattice bundle and the offset bundle. It is calculated based on the correction table prepared in advance by the detailed calculation of the multi-assembly according to the combination with the combination, the diffusion node method using the assembly discontinuity factor, or the analytical diffusion model. It has an excellent effect that the power distribution, the maximum linear power density, the limit power ratio, the neutron detector count in the reactor, the detector irradiation amount, the control rod irradiation amount, and the like can be accurately calculated.

In addition, for the channel box deformation due to the fast neutron irradiation, the fast neutron irradiation amount on each surface of the channel box is integrated by core power distribution calculation,
Since the amount of channel box deformation caused by neutron irradiation can be predicted and evaluated based on the irradiation growth model, it becomes possible to calculate the core core characteristics in consideration of the channel box deformation. Further, the correction calculation of the core core characteristics in consideration of the channel box deformation is performed based on the correction table or the diffusion model prepared in advance by the multi-aggregate detailed calculation based on the channel box deformation amount of the target aggregate and the adjacent aggregate. Since the effect is calculated after correcting for the hysteresis effect,
It has an excellent effect that the critical eigenvalue of the core, the power distribution in the furnace, the maximum linear power density, the critical power ratio, the neutron detector count in the reactor, the detector irradiation amount, the control rod irradiation amount, and the like can be accurately calculated.

[Brief description of the drawings]

FIG. 1 is a block diagram showing a first embodiment of the present invention.

FIG. 2 is an explanatory diagram showing a combination pattern of a D lattice bundle and an offset bundle.

FIG. 3 is a characteristic diagram showing an example of burnup dependence of a local peaking coefficient in an offset bundle adjacent system.

FIG. 4 is a characteristic diagram showing a relationship between an offset amount of a channel box and a neutron flux discontinuity factor of an aggregate.

FIG. 5 is an explanatory diagram showing the effect of local output distribution calculation according to the first embodiment of the present invention.

FIG. 6 is a block diagram showing a second embodiment of the present invention.

FIG. 7 is an explanatory diagram showing an evaluation model of the amount of bending in the axial direction of the channel box due to irradiation growth.

FIG. 8 is a characteristic diagram showing an example of an irradiation growth function by fast neutron irradiation.

FIG. 9 is an explanatory diagram in a case where an axial bending amount of a channel box due to irradiation growth is obtained by a finite element method.

FIG. 10 is a characteristic diagram showing an example of an irradiation growth function depending on burnup.

FIG. 11 is an explanatory diagram illustrating a relationship between a channel deformation amount and a water gap deformation amount according to the second embodiment of the present invention.

FIG. 12 shows a water gap width deformation amount (δx = δ) of an aggregate of interest.
FIG. 8 is a characteristic diagram showing an example of a change in the maximum fuel rod local output with respect to y and δy = 0).

FIG. 13 shows a water gap width deformation amount (dx = d) of an adjacent aggregate.
FIG. 8 is a characteristic diagram showing an example of a change in the maximum fuel rod local output with respect to y and dy = 0).

FIG. 14 is a characteristic diagram showing an example of a change in an aggregate neutron flux discontinuity factor with respect to a water gap width deformation amount of an aggregate of interest.

FIG. 15 is a partially cutaway perspective view of a fuel assembly of a boiling water reactor.

FIG. 16 is an explanatory diagram of a positional relationship of a water gap in a fuel assembly of a boiling water reactor.

FIG. 17 is an explanatory diagram of a D lattice bundle.

FIG. 18 is an explanatory diagram showing a state in which two bodies around a control rod are replaced with offset bundles.

FIG. 19 is an explanatory diagram showing an example of a change in a local output distribution in an aggregate due to adjacent offset bundles.

FIG. 20 is an explanatory diagram showing an axial bending of a channel box due to fast neutron irradiation.

FIG. 21 is an explanatory diagram showing a positional relationship of a water gap due to deformation of a channel box.

[Explanation of symbols]

 DESCRIPTION OF SYMBOLS 1 Fuel assembly 2 Channel box 3 Fuel rod 4 Water gap 5 Control rod 6 Instrumentation pipe 11 Assembly average nuclear constant calculation means 12 Furnace power distribution calculation means 13 Local power distribution calculation means 14 Line power calculation means 15 Critical power ratio Calculation means 16 In-core neutron flux meter counting value calculation means 17 In-core power distribution correction means 18 Control rod irradiation amount calculation means 19 In-core power distribution calculation means 20 Channel box irradiation amount calculation means 21 Channel box deformation amount calculation means 22 Furnace Internal power distribution recalculation means 23 Linear power density calculation means 24 Limit power ratio calculation means 25 In-core neutron flux meter counting value calculation means 26 Control rod irradiation dose calculation means

Claims (44)

[Claims] 1. A core performance calculation of a boiling water reactor in which a D lattice bundle having a larger water gap width on the control rod side than other sides and an offset bundle in which the channel center of the D lattice bundle is eccentric are mixed. The apparatus calculates an aggregate average nuclear constant using a correction amount for an aggregate average nuclear constant or an infinite lattice average nuclear constant that is evaluated in advance by a multi-aggregate detailed calculation according to a combination of an adjacent D lattice bundle and an offset bundle. Based on the core diffusion calculated using the core constant calculated by the aggregate average nuclear constant calculating means, and the core eigenvalue, neutron flux distribution, and power output in the core to calculate the power distribution. A core performance calculation device comprising distribution calculation means. 2. A core performance calculation of a boiling water reactor in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is eccentric is mixed. In the apparatus, in-core power distribution calculation means for calculating the critical eigenvalue, neutron flux distribution and power distribution of the core based on the core diffusion calculation, and multi-aggregate detailed calculation according to the combination of adjacent D lattice bundle and offset bundle The local power distribution calculating means for calculating the fuel rod local output using the correction amount for the fuel rod local output or the infinite lattice fuel rod local power in the fuel assembly evaluated in advance, and the in-furnace power distribution calculating means The linear power density of the assembly is calculated using the average output of the cross section of the assembly and the local output of the fuel rod calculated by the local power distribution calculating means. A core performance calculation device comprising: a linear power density calculation unit. 3. A core performance calculation of a boiling water reactor in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is eccentric is mixed. In the apparatus, in-core power distribution calculation means for calculating the critical eigenvalue, neutron flux distribution and power distribution of the core based on the core diffusion calculation, and multi-aggregate detailed calculation according to the combination of adjacent D lattice bundle and offset bundle The fuel rod R-factor is calculated by using the correction amount for the fuel rod R-factor or the infinite lattice fuel rod R-factor in the fuel assembly evaluated in advance, and the average power of the assembly calculated by the in-furnace power distribution calculating means and the calculation. And a limit power ratio calculating means for calculating a limit power ratio of the assembly using the obtained fuel rod R factor. 4. A local output of a fuel rod in a fuel assembly, which has been evaluated in advance by a detailed multi-assembly calculation for a combination of a D lattice bundle and an offset bundle in which a thermal limit value such as a maximum linear power density or a critical power ratio is the strictest. ,
The linear power density and the critical power ratio of the core are corrected and calculated by the core diffusion calculation using the correction amount for the fuel rod R factor, respectively.
A core performance calculation device according to item 1. 5. A core performance calculation of a boiling water reactor in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is eccentric is mixed. In the apparatus, in-core power distribution calculation means for calculating the critical eigenvalue, neutron flux distribution and power distribution of the core based on the core diffusion calculation, and multi-aggregate detailed calculation according to the combination of adjacent D lattice bundle and offset bundle The in-core measuring device position neutron flux calculated using the correction amount for the in-core measuring device position neutron flux or the infinite lattice measuring device position neutron flux in advance and the aggregate calculated by the in-core power distribution calculating means And a means for calculating a neutron flux meter in the reactor based on the average output of the cross section and calculating the irradiation dose. Core performance calculator. 6. A core performance calculation device for a boiling water reactor in which a D lattice bundle having a larger water gap width than the other side of the control rod example and an offset bundle in which the channel center of the D lattice bundle is eccentric is mixed. In the core diffusion calculation, based on the core eigenvalue, neutron flux distribution, power distribution calculation means for calculating the power distribution, and multi-aggregate detailed calculation according to the combination of adjacent D lattice bundle and offset bundle The control rod position neutron flux calculated using the correction amount for the control rod position neutron flux or the infinite lattice control rod position neutron flux evaluated in advance and the control rod position neutron flux based on the aggregate cross-sectional average power calculated by the in-furnace power distribution calculation means are used to irradiate the control rod. And a control rod irradiation amount calculating means for calculating the amount. 7. The core performance calculation according to claim 2, wherein the aggregate average power calculated and corrected by said aggregate average nuclear constant calculation means is used as the aggregate average output. apparatus. 8. A core performance calculation of a boiling water reactor in which a D lattice bundle having a larger water gap width on the control rod side than the other side and an offset bundle in which the channel center of the D lattice bundle is eccentric is mixed. A thermal neutron flux change calculating means for calculating a change of a thermal neutron flux distribution in the assembly with respect to a change amount of the water gap width of the focused assembly from the infinite lattice system based on an analytical model for a diffusion equation; A linear power density calculating means for correcting a fuel rod local output using the thermal neutron flux distribution change calculated by the neutron flux change calculating means to calculate a linear power density of the assembly; and a thermal neutron flux change calculating means. Critical power ratio calculating means for correcting the fuel rod R factor using the calculated thermal neutron flux distribution change to calculate the critical power ratio of the assembly, and the thermal neutron flux change calculating means. In-core neutron flux meter counting value calculating means for correcting in-core neutron flux in the reactor using the calculated thermal neutron flux distribution change and calculating in-core neutron flux meter counting and irradiation amount, the thermal neutron A control rod irradiation amount calculating means for calculating a control rod irradiation amount by correcting a control rod position neutron flux by using the thermal neutron flux distribution change calculated by the flux change calculating means; . 9. A core power distribution calculation means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor, and a neutron flux discontinuity factor. Calculates the homogeneous thermal neutron flux distribution in the node where each aggregate is homogenized based on the diffusion theory using, and evaluated in the infinite lattice calculation, "The homogeneous thermal neutron flux of the inhomogeneous thermal neutron flux at the aggregate boundary Of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the aggregate by multiplying the factor by multiplying the ratio by Having a non-homogeneous thermal neutron flux calculation means for calculating the non-homogeneous thermal neutron flux distribution, and correcting the aggregate average nuclear constant using the non-homogeneous neutron flux distribution by the non-homogeneous thermal neutron flux calculation means to correct the core eigenvalue , A core performance calculation device characterized in that a neutron flux distribution and a power distribution in a reactor are calculated. 10. A core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor,
Based on the diffusion theory using neutron flux discontinuity factor, we calculated the homogeneous thermal neutron flux distribution in the nodes where each aggregate was homogenized, and evaluated the inhomogeneous thermal neutron flux at the boundary of the aggregate evaluated in the infinite lattice calculation. Multiplied by the ratio of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the assembly, and the correction factor for the transient component due to thermal neutron diffusion of the homogeneous thermal neutron flux distribution. A non-homogeneous neutron flux calculation means for calculating a non-homogeneous thermal neutron flux distribution in the assembly, and an aggregate cross-sectional average power calculated by the in-core power distribution calculation means and a non-homogeneous thermal neutron flux calculation means. A core power calculating device for calculating a linear power density of a fuel assembly by using a fuel rod local power calculated from a homogeneous neutron flux distribution; 11. A core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor,
Based on the diffusion theory using neutron flux discontinuity factor, we calculated the homogeneous thermal neutron flux distribution in the nodes where each aggregate was homogenized, and evaluated the inhomogeneous thermal neutron flux at the boundary of the aggregate evaluated in the infinite lattice calculation. Multiplied by the ratio of the homogeneous thermal neutron flux to the inhomogeneous thermal neutron flux in the assembly, and the correction factor for the transient component due to thermal neutron diffusion of the homogeneous thermal neutron flux distribution. A non-homogeneous neutron flux calculation means for calculating a non-homogeneous thermal neutron flux distribution in the assembly, and an aggregate cross-sectional average power calculated by the in-core power distribution calculation means and a non-homogeneous thermal neutron flux calculation means. A linear power density calculating means for calculating a linear power density of the fuel assembly by using the fuel rod local power calculated from the homogeneous neutron flux distribution; and And a limit power ratio calculating means for calculating a limit power ratio of the fuel assembly from a fuel rod R factor calculated by using the fuel rod local factor and the fuel rod local output by the linear power density calculating means. Core performance calculator. 12. A core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor,
Node boundary value calculation means for calculating neutron flux and neutron flow at the node boundary surface where each assembly is homogenized, based on the diffusion theory calculation using the neutron flux discontinuity factor, and calculation by the in-core power distribution calculation means The line of the assembly by the infinite lattice fuel rod local output corrected using the calculated cross-sectional average output of the assembly and the sensitivity coefficient of the local output of the fuel rod to the neutron flux and neutron flow on each surface of the assembly boundary by the node boundary value calculation means A core power calculation device for calculating a power density. 13. A core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor,
Node boundary value calculation means for calculating neutron flux and neutron flow at the node boundary surface where each assembly is homogenized, based on the diffusion theory calculation using the neutron flux discontinuity factor, and calculation by the in-core power distribution calculation means The line of the assembly by the infinite lattice fuel rod local output corrected by using the adjusted cross-sectional average output and the sensitivity coefficient of the fuel rod local output to the neutron flux and neutron flow on each surface of the assembly boundary by the node boundary value calculation means Linear power density calculating means for calculating the power density, and a fuel rod R factor calculated by using the assembly cross-sectional average power calculated by the in-furnace power distribution calculating means and the fuel rod local power by the linear power density calculating means And a limit power ratio calculating means for calculating a limit power ratio of the fuel assembly. 14. A core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution, and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor,
Node boundary value calculation means for calculating neutron flux and neutron flow at the node boundary surface where each assembly is homogenized, based on the diffusion theory calculation using the neutron flux discontinuity factor, and calculation by the in-core power distribution calculation means Limit output ratio of the aggregate by the infinite lattice R factor corrected using the calculated average output of the aggregate and the sensitivity factor of the fuel rod R factor to the neutron flux and neutron flow on each surface of the aggregate boundary by the node boundary value calculation means. And a limit power ratio calculating means for calculating the core power ratio. 15. A core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor,
Node boundary value calculation means for calculating neutron flux and neutron flow at the node boundary surface where each assembly is homogenized, based on the diffusion theory calculation using the neutron flux discontinuity factor, and calculation by the in-core power distribution calculation means From the neutron flux at the infinite lattice measuring instrument position corrected by using the sensitivity coefficient of the neutron flux at the measuring instrument position in the reactor with respect to the neutron flux and the neutron flux on each surface of the aggregate boundary by the calculated aggregate cross section average output and the node boundary value calculating means An in-core neutron flux meter counting means for calculating a neutron flux meter and an irradiation dose in the reactor. 16. A core power distribution calculating means for calculating a critical eigenvalue, a neutron flux distribution and a power distribution of a core based on a core diffusion calculation in a core performance calculation of a boiling water reactor,
Node boundary value calculation means for calculating neutron flux and neutron flow at the node boundary surface where each assembly is homogenized, based on the diffusion theory calculation using the neutron flux discontinuity factor, and calculation by the in-core power distribution calculation means Control rod position for the neutron flux and neutron flux on the neutron flux and neutron flow on each side of the aggregate boundary by the calculated aggregate cross section average output and the node boundary value calculation means. A core performance calculation device, comprising: a control rod irradiation amount calculating means for calculating an irradiation amount. 17. The in-furnace power distribution calculating means according to claim 1,
Calculate the fast neutron flux distribution by group diffusion calculation and calculate the thermal neutron flux and thermal neutron flow using the obtained fast neutron flux and the spectral index that is the ratio of the thermal neutron flux to the fast neutron flux in the infinite lattice calculation. 17. The core performance calculation device according to claim 9, wherein: 18. The in-core power distribution by adapting the calculated value of the in-core neutron flux meter corrected for the adjacent effect between the D lattice bundle and the offset bundle to the actually measured value of the in-core neutron flux meter. 2. A furnace power distribution correcting means for correcting a calculated value is provided.
8. The core performance calculation device according to 7. 19. Using the burn-up distribution in the fuel assembly and the spectrum history distribution, which is the burn-up average value of the ratio of the thermal neutron flux to the fast neutron flux, by the hysteresis effect associated with the adjacency between the D lattice bundle and the offset bundle. 19. The core performance calculation device according to claim 9, wherein a correction amount of the assembly nuclear constant, the fuel rod local power distribution, and the fuel rod R-factor distribution is calculated. 20. A core performance calculation device for a boiling water reactor, a reactor power distribution calculating means for calculating a neutron flux and a power distribution in the core, and a channel for calculating a fast neutron irradiation amount on each surface of the channel box. Box irradiation amount calculating means, and channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means And calculating an aggregate nuclear constant and a neutron flux discontinuity factor in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculation means, and calculating the corrected aggregate nuclear constant and neutron. Reactor that calculates critical eigenvalue and power distribution of core using flux discontinuity factor An internal power distribution recalculating means, comprising: a core performance calculating device. 21. A core performance calculation device for a boiling water reactor, a reactor power distribution calculation means for calculating a neutron flux and a power distribution in the core, and a channel for calculating a fast neutron irradiation dose on each surface of the channel box. Box irradiation amount calculating means, and channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means And calculating the local output of the fuel rod in the fuel assembly in which the channel box deformation effect is corrected by using the channel box deformation amount evaluated by the channel box deformation amount calculating means, thereby giving the maximum local output peaking coefficient of the assembly. And cross-sectional average power of the aggregate by the core power distribution calculation means And a linear power density calculating means for calculating a linear power density of each fuel assembly cross section by using the above. 22. An apparatus for calculating a core performance of a boiling water reactor, comprising: a reactor power distribution calculating means for calculating a neutron flux and a power distribution in the core; and a channel for calculating a fast neutron irradiation amount on each surface of the channel box. Box irradiation amount calculating means, and channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means And calculating a fuel rod R factor representing a fuel rod output pattern in the assembly in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculating means. And the aggregate average power by the core power distribution calculation means, A core performance calculation device, comprising: a limit power ratio calculation means for calculating a limit power ratio, which is a monitoring index of burnout of a fuel rod of a fuel assembly. 23. An apparatus for calculating a core performance of a boiling water reactor, a means for calculating a neutron flux and a power distribution in the core, and a channel for calculating a fast neutron irradiation dose on each surface of the channel box. Box irradiation amount calculating means, and channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means And an in-core neutron flux for calculating the in-core neutron flux meter count value and the in-core neutron flux meter irradiation amount in which the channel box deformation effect is corrected using the channel box deformation amount evaluated by the channel box deformation amount calculation means. A core performance calculation device, comprising: a measurement device measurement value calculation unit. 24. An apparatus for calculating the core performance of a boiling water reactor, comprising: a reactor power distribution calculating means for calculating a neutron flux and a power distribution in the core; and a channel for calculating a fast neutron irradiation amount on each surface of the channel box. Box irradiation amount calculating means, and channel box deformation amount calculating means for evaluating the amount of deformation of the channel box due to neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface evaluated by the channel box irradiation amount calculating means And control rod irradiation amount calculating means for calculating a control rod irradiation amount using a control rod position neutron flux in which a channel box deformation effect is corrected by the channel box deformation amount evaluated by the channel box deformation amount calculating means. Core performance calculator. 25. The channel box deformation amount calculation means calculates the irradiation growth amount on the opposing surface of the channel box as a function of the fast neutron irradiation amount on each surface, and calculates the channel box from the difference in the irradiation growth amount on the opposing surface. 25. The core performance calculation apparatus according to claim 20, wherein the amount of deformation is calculated. 26. An apparatus for calculating core performance of a boiling water reactor, comprising: in-core power calculating means for calculating neutron flux and power distribution in the core; and burn-up distribution calculating means for calculating burn-up distribution in the core. Using the burn-up of each surface of the channel box evaluated by the burn-up distribution calculating means, the amount of deformation of the channel box due to neutron irradiation and the fact that the irradiation growth amount of the channel box corresponds one-to-one to the burn-up Channel box deformation amount calculating means evaluated using an irradiation growth model, and an aggregate nuclear constant and a neutron flux discontinuity factor corrected for a channel box deformation effect using the channel box deformation amount evaluated by the channel box deformation amount calculating means Is calculated, and the corrected core nuclear constant and neutron flux discontinuity factor are used to calculate the critical eigenvalue of the core and A core performance calculation device comprising: a power distribution recalculation means for calculating a power distribution in a reactor. 27. An apparatus for calculating the core performance of a boiling water reactor, comprising: a power distribution calculation means for calculating a neutron flux and a power distribution in the core; and a burnup distribution calculation means for calculating a burnup distribution in the core. And, using the burnup of each surface of the channel box evaluated by the burnup distribution calculation means,
Channel box deformation amount calculating means for evaluating the deformation amount of the channel box due to neutron irradiation using an irradiation growth model using that the irradiation growth amount of the channel box corresponds one-to-one to the burnup, and the channel box deformation amount calculation Means for calculating the local power of the fuel rods in the fuel assembly with the channel box deformation effect corrected using the channel box deformation amount evaluated by the means, and the maximum local power peaking coefficient of the assembly and the core power distribution calculation means provided by the calculation. A core power density calculating means for calculating a linear power density of each fuel assembly cross section by using the average cross section power of the fuel assembly. 28. An apparatus for calculating the core performance of a boiling water reactor, comprising: an in-core power distribution calculating means for calculating a neutron flux and a power distribution in the core; and a burn-up distribution calculating means for calculating a burn-up distribution in the core. And, using the burnup of each surface of the channel box evaluated by the burnup distribution calculation means,
Channel box deformation amount calculating means for evaluating the deformation amount of the channel box due to neutron irradiation using an irradiation growth model using that the irradiation growth amount of the channel box corresponds one-to-one to the burnup, and the channel box deformation amount calculation Means for calculating a fuel rod R factor representing a fuel rod output pattern in the assembly in which the channel box deformation effect is corrected by using the channel box deformation amount evaluated by the means, and the maximum R factor of the assembly and the in-core power distribution calculating means. And a limit power ratio calculating means for calculating a limit power ratio, which is an index for monitoring burnout of fuel rods of each fuel assembly, using the average power of the fuel assemblies. 29. An apparatus for calculating a core performance of a boiling water reactor, comprising: a core power distribution calculating means for calculating a neutron flux and a power distribution in the core; and a burnup distribution calculating means for calculating a burnup distribution in the core. And, using the burnup of each surface of the channel box evaluated by the burnup distribution calculation means,
Channel box deformation amount calculating means for evaluating the deformation amount of the channel box due to neutron irradiation using an irradiation growth model using that the irradiation growth amount of the channel box corresponds one-to-one to the burnup, and the channel box deformation amount calculation Using the channel box deformation amount evaluated by the means, the in-core neutron flux meter count value and the in-core neutron flux meter count value calculation means for calculating the in-core neutron flux meter count and the in-core neutron flux meter irradiation amount in which the channel box deformation effect is corrected A core performance calculation device comprising: 30. An apparatus for calculating a core performance of a boiling water reactor, comprising: a core power distribution calculating means for calculating a neutron flux and a power distribution in the core; and a burnup distribution calculating means for calculating a burnup distribution in the core. And, using the burnup of each surface of the channel box evaluated by the burnup distribution calculation means,
Channel box deformation amount calculating means for evaluating the deformation amount of the channel box due to neutron irradiation using an irradiation growth model using that the irradiation growth amount of the channel box corresponds one-to-one to the burnup, and the channel box deformation amount calculation And a control rod irradiation amount calculating means for calculating a control rod irradiation amount using a control rod position neutron flux in which a channel box deformation effect is corrected by the channel box deformation amount evaluated by the means. 31. The channel box deformation amount calculation means calculates the irradiation growth amount on the opposite surface of the channel box as a function of the burnup of each surface of the channel box, and calculates the channel box from the difference in the irradiation growth amount on the opposite surface. 31. The core performance calculation apparatus according to claim 27, wherein the amount of deformation is calculated. 32. The in-furnace power distribution recalculating means includes: an aggregate average nuclear constant, a fuel rod local power distribution, a fuel rod local power distribution, and a channel box deformation amount of an aggregate of interest and an adjacent aggregate evaluated in advance by multi-aggregate detailed calculation Change of nuclear characteristics of fuel assembly such as rod maximum local output peaking coefficient, fuel rod R-factor distribution, fuel rod maximum R-factor, neutron flux in the reactor neutron flux detector, neutron flux in the control rod, etc. It is characterized in that it is arranged as a table or fitting coefficient for the water gap width of each surface, and by referring to using the water gap width obtained from the channel box deformation, the fuel assembly core characteristics are calculated. 32. The core performance calculation device according to claim 20. 33. The in-furnace power distribution recalculating means calculates a change in the neutron flux discontinuity factor of the aggregate with respect to the amount of channel box deformation estimated in advance by the single aggregate detailed calculation, and calculates a change of each surface of the channel box of the aggregate of interest. By arranging as a table or fitting coefficient for the water gap width and referring to using the water gap width obtained from the channel box deformation, the neutron flux distribution in the aggregate is calculated based on the diffusion model, and this neutron flux is calculated. Using the distribution, average nuclear constant of the assembly, fuel rod local power distribution, fuel rod maximum local power peaking coefficient, fuel rod R factor distribution, fuel rod maximum R factor, neutron flux in the reactor neutron flux measuring device, neutron flux in the control rod position, neutrons in the control rod position 32. The core performance calculation device according to claim 20, wherein the core characteristics of the fuel assembly such as a bundle are calculated. . 34. The in-furnace power distribution recalculating means calculates a change of the neutron flux discontinuity factor of the aggregate with respect to the amount of channel box deformation evaluated in advance by the single aggregate detailed calculation, for each surface of the channel box of the aggregate of interest. Calculate the neutron flux and neutron flow on each surface of the assembly boundary based on the diffusion model by organizing a table or fitting coefficient for the water gap width and referencing using the water gap width obtained from the channel box deformation. The aggregate average nuclear constant for the neutron flux and neutron flow on each surface evaluated in advance, the fuel rod local power distribution, the fuel rod maximum local power peaking coefficient, the fuel-like R factor distribution, the fuel rod maximum R factor, the neutron in the reactor Using the sensitivity coefficients of the nuclear characteristics of the fuel assembly such as the neutron flux at the flux detector and the neutron flux at the control rod, 32. The core performance calculation device according to claim 20, wherein the calculation is performed. 35. The in-furnace power distribution recalculating means calculates a change in a thermal neutron flux distribution in the aggregate with respect to a channel box deformation amount of the aggregate of interest and the adjacent aggregate based on an analytical model for a diffusion equation, Using this thermal neutron flux distribution, the aggregate average nuclear constant, fuel rod local power distribution,
Calculation of fuel assembly nuclear properties such as maximum local power peaking coefficient of fuel rod, distribution of fuel rod R-factor, maximum fuel rod R-factor, neutron flux in the reactor neutron flux detector, neutron flux in the control rod, etc. Claims 20 to 3 characterized by the above-mentioned.
2. The core performance calculation device according to 1. 36. The in-core power distribution calculation by adapting the calculated value of the in-core neutron flux measurement device in which the deformation effect of the channel box due to the neutron irradiation is corrected to the actually measured value of the in-core neutron flux measurement device. 32. The core performance calculation device according to claim 20, wherein the value is corrected. 37. The channel box deformation effect is corrected according to claim 20, as the fuel assembly average output used for calculating the linear power density, the critical power ratio, the in-reactor neutron flux meter counting and irradiation amount, and the control rod irradiation amount. 25. The core performance calculation device according to claim 21, wherein the average fuel assembly output is used. 38. The channel box deformation effect is corrected according to claim 26, as the average output of the fuel assembly used for calculating the linear power density, the critical power ratio, the in-reactor neutron flux meter counting and irradiation amount, and the control rod irradiation amount. 31. The fuel assembly average output is used.
A core performance calculation device according to item 1. 39. The combustion of the deformation of the channel box using the burn-up distribution in the fuel assembly considering the deformation effect of the channel box and the spectrum history distribution which is the burn-up average value of the ratio of the thermal neutron flux to the fast neutron flux. The correction amount of the aggregate nuclear constant, the fuel rod local output distribution, and the fuel rod R factor distribution due to the accompanying hysteresis effect is calculated. A core performance calculator according to the above description. 40. A fast neutron flux distribution is calculated by a modified one-group diffusion calculation, and a thermal neutron flux is calculated using a spectrum index which is a ratio of the obtained fast neutron flux to a fast neutron flux in an infinite lattice system. 35. The core performance calculation device according to claim 33, wherein the core neutron flow and the thermal neutron flow are calculated. 41. When evaluating the amount of deformation of the channel box due to the neutron irradiation using an irradiation growth model,
26. The reactor core performance calculating apparatus according to claim 25, wherein a ratio of neutrons of all neutrons having a neutron energy of 1 MeV or more to all fast neutrons is calculated as a table of a burnup and a void ratio or a fitting coefficient. 42. The channel box deformation amount calculating means predicts and calculates a channel box deformation amount by using a past measured value of the channel box deformation amount. 3
2. The core performance calculation device according to 1. 43. The amount of deformation of the channel box due to the neutron irradiation is analytically obtained assuming that the axial bending of the channel box has a uniform curvature according to the difference in the amount of irradiation growth on the opposing surface of the channel box. 32. The method according to claim 25, wherein:
A core performance calculation device according to item 1. 44. The amount of deformation of the channel box due to the neutron irradiation is calculated by dividing the channel box into axial nodes, obtaining the length of the surface opposite to the channel after irradiation growth for each node, and stacking the resultant. 32. The core performance calculating device according to claim 25, wherein a distance between a straight line connecting upper and lower ends and each node is obtained as a bending amount.