JP4008926B2 - Core performance calculator - Google Patents

Description

  The present invention relates to a core performance calculation apparatus in a boiling water reactor.

  The core of a boiling water reactor is configured after a plurality of fuel assemblies (bundles) are inserted. FIG. 15 is a partially cutaway perspective view of the fuel assembly 1. As shown in FIG. 15, the fuel assembly 1 is configured by bundling a plurality of fuel rods 3 in a channel box 2. In the core, cooling water enters from the lower part of the core, so that in the channel box 2, the cooling water flows from the lower direction to the upper direction. 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 diagram of the arrangement of the fuel assemblies 1 in the core. As shown in FIG. 16, a water gap 4 for flowing non-boiling cooling water exists between channel boxes 2 a to 2 d of four adjacent fuel assemblies 1. Control rods 5 are provided as necessary at the positions of the water gaps 4 at the corners sandwiched between the channel boxes 2a to 2d of the four fuel assemblies 1, and the water gaps 4 at the corners at the diagonal positions are provided. Is provided with an instrumentation tube 6 for the in-core neutron flux measuring instrument.

  Here, as shown in FIG. 17, the initial boiling water reactor is a so-called D lattice 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. A fuel core is used. Such a D-lattice fuel assembly has a large neutron moderation effect in a wide water gap LW (hereinafter referred to as a wide gap), so that the thermal neutron flux on the wide gap LW side is narrower than that on the narrow water gap LN (hereinafter referred to as narrow gap) side. As a result, the thermal neutron flux distribution in the fuel assembly 1 becomes asymmetric. For this reason, in order to flatten the fuel rod local power distribution, it is necessary to reduce the enrichment of the fuel rod 3 on the wide gap LW side and increase the enrichment on the narrow gap LN side.

  Thus, if the enrichment distribution is not symmetrical, for example, when the channel box shakes during an earthquake, the thermal neutron flux distribution may shift to the highly enriched side due to the change in the water gap width. In such a case, reactivity may be applied, leading to scram. In order to avoid this, it is conceivable to make the widths of the wide gap LW and the narrow gap LN closer 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 to the control rod 5 side within a range not interfering with the control rod 5. The D lattice bundle in which the center of the channel box 2 is eccentric to the control rod 5 side is called an offset bundle.

  This offsetting is realized by replacing the D lattice bundle with the offset bundle step by step for 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 boxes 2b and 2c of the fuel assembly 1 are offset.

  In the core performance calculation in the nuclear reactor, the neutron flux distribution and power distribution in the core are calculated based on the three-dimensional nuclear thermal-hydraulic coupled physical model of the core. The fuel rod power normalized by the cross section of the assembly is called local power, and the thermal limit value of the core is calculated from the average power of the cross section of the assembly and the local power of the fuel rod. The maximum line power density of the assembly is defined as the product of the maximum fuel rod local power peaking coefficient of the assembly cross section and the assembly cross-section average output. Further, the limit 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 local power distribution around each fuel rod, the assembly average power, and the channel flow rate. .

  The calculation method of these thermal limit values is described in, for example, the document “Three-dimensional BWR core simulator,” J.A. Wooly, Licensing topic a1 report, NEDO-20953, 1976, General E1ectric Company.

  In order to improve the accuracy of the three-dimensional nuclear thermal-hydraulic coupled physical model of the core, for example, the document "TARMS: An on-Line Boiling Water Reactor Management System Based on Core Physics Simulator," M. Tsuiki et al., Proceedings of As described in a topical meeting on Advances in Reactor Computations, Salt Lake City, 1983, the measured and calculated values of the neutron flux detector in the core are used to correct the neutron flux distribution in the core calculation. Improvements have been made to the accuracy.

  Furthermore, 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, in order to monitor the life of the control rod, the amount of control rod irradiation is integrated.

  In the core three-dimensional calculation, the core average eigenvalue for the homogenized node of each assembly is used, and the core core core eigenvalues, the in-core neutral flux distribution, and the in-core power distribution are determined by the coarse mesh node method based on the neutron diffusion theory. Calculate The aggregate average nuclear constant is generally calculated by 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 aggregate of interest and the adjacent aggregate is the same, so the infinite lattice calculation using the reflection boundary condition is a good approximation.

  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 approximation of the reflection boundary condition in the infinite lattice system when the respective node average nuclear constants are calculated becomes worse. This is because the distribution of the thermal neutron flux locally changes in the vicinity of the water gap due to the change in the water gap width, resulting in a change in the aggregate nuclear characteristics. It should be noted that the fast neutron flux hardly changes, and it can be said that the change in the aggregate nucleus characteristics is almost caused by the change in the vicinity of the water gap of the thermal neutron flux. This change occurs in both adjacent D-lattice bundles and offset bundles.

  If the offset amount is large, the effect on the core characteristics will be large.In the core performance calculation, the core critical eigenvalue, the reactor power distribution, the fuel rod power distribution in the reactor, the thermal limit value, the neutron detector count value, In order to monitor the performance of the core, it is necessary to consider the thermal neutron flux change at the control rod position.

  As shown in FIG. 18, in the example of the nuclear property change by the offset bundle, consider a case where two offset bundles, each having a channel box center offset by 2 mm toward the control rod, are loaded around the control rod. FIG. 19 shows the characteristics of the local power distribution in the assembly in the D lattice bundle in this case compared with the fuel rod local power distribution of the infinite lattice system. In FIG. 19, the upper part shows a mixed system, the middle part shows an infinite lattice system, and the lower part shows the difference.

  In this case, the local output distribution of the reference solution was evaluated by multi-aggregate inhomogeneous calculation. Adjacent offset bundles will substantially reduce the wide gap of the D lattice bundle and increase the narrow gap. According to the reference solution, the maximum local peaking coefficient of the D-lattice bundle changes by about 10% due to the adjacency of the offset bundle, and the importance of core monitoring in consideration of the adjacency effect with the offset bundle is shown. Has 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 made of, for example, a zirconium alloy. The upper and lower ends of the fuel rod are supported by the upper tie plate and the 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 amount of fast neutron irradiation between the surfaces of the opposing channel box 2, as shown in FIG. 20, a difference occurs in the amount of irradiation growth in the axial direction of the channel box 2, so that 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.

  Thus, when the channel box 2 is bent, as shown in FIG. 21, the width of the water gap 4 of the fuel assembly 1 changes from the normal width. Therefore, mechanically, there is an effect that the control rod 5 is prevented from being inserted into the water gap 4.

  When the channel box of the target assembly and the adjacent assembly is bent due to irradiation growth, the width of the water gap between adjacent fuel assemblies changes from the normal width, which changes the nuclear characteristics of the fuel assembly. Arise. 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 is bent due to irradiation growth, the width of the water gap between the adjacent fuel assemblies changes from the normal width. When the gap width changes, the neutron moderation effect and absorption effect are changed, and the thermal neutron flux distribution in the vicinity of the water gap changes, resulting in a change in the nuclear characteristics of the fuel assembly.

  It should be noted that the fast neutron flux hardly changes, and it can be said that the change in the aggregate nucleus characteristics is almost caused by the change in the vicinity of the water gap of the thermal neutron flux. When the channel box deformation is large, the effect on the core characteristics 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, neutron detection In order to monitor the performance of the core, it is necessary to take into account the count value of the reactor and the change in thermal neutron flux at the control rod position.

  Conventionally, the operation management of the core that predicts the deformation of the channel box 2 due to neutron irradiation and avoids the interference between the control rod 5 and the channel box 2 is described in Patent Document 1, Patent Document 2, and Patent Document 3. Is shown in

  However, these documents do not consider the influence on the core performance calculation due to the deformation of the channel box. An example of changes in nuclear characteristics due to channel box deformation will be described. For example, as shown in FIG. 21, due to the deformation of the channel box 2, the water gap 4 on the control rod 5 side of the fuel assembly 1 of interest now decreases by 2 mm, and the water gap 4 on the instrumentation tube 6 side for the neutron detector Is increased by 2 mm, and other adjacent assemblies 1 are not deformed. 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 the same characteristics as FIG. 19 when the offset bundle is employed.

  In this case, the local output distribution of the reference solution was evaluated by multi-aggregate inhomogeneous calculation. According to this, the aggregate maximum local peaking coefficient is changed by about 10% due to the channel box deformation, which shows the importance of monitoring the core in consideration of the channel box deformation.

Here, Patent Document 4 shows that 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 design normal position in the core performance calculation. .
Japanese Patent Laid-Open No. 2-176497 JP-A-2-201291 Japanese Patent Laid-Open No. 4-204084 JP-A-4-110698

  In Patent Document 4, four fuel assemblies surrounding a control rod in a core or a neutron detector in a reactor are defined as one unit cell, and a change in fuel rod output due to misalignment of the assembly in each unit cell is calculated. Yes.

  In general, the fuel rod output change in the target assembly in the unit cell depends on the combination of the amount of misalignment and the direction of misalignment of the four fuel assemblies in the unit cell. Only the calculation based on the amount of misalignment of the fuel assembly is described, and the evaluation method is not clear.

  Moreover, in the thing of patent document 4, the influence of the position shift of the assembly | attachment adjacent to a focused assembly from the outside of the focused unit cell is disregarded, and there existed a problem that a fuel rod output change could not be evaluated accurately. Furthermore, there is no specific method for evaluating the limit output ratio from the change in fuel rod output due to the misalignment of the assembly.

  Further, in Patent Document 4, there is no influence on the critical eigenvalue of the core, the power distribution in the reactor core, the count value of the neutron detector in the reactor, the detector irradiation amount, the control rod irradiation amount, etc. due to the displacement of the fuel assembly. Not considered. In particular, the count of the neutron flux detector in the reactor is sensitive to changes in the water gap width, and in order to learn the error from the measured value of the neutron flux detector and correct the reactor power distribution, the calculated value of the detector must be accurate. It is necessary to ask well.

  In addition, if combustion proceeds with the fuel assembly misaligned, the combustion history effect and burnup distribution effect of neutron spectrum changes in the fuel assembly due to changes in the water gap width will accumulate, and the assembly will be misaligned instantaneously The difference with the effect of is produced. This history effect generally occurs in a direction that cancels out the instantaneous effect. However, in Patent Document 4, this history effect is not considered at all.

  Regarding the deformation of the channel box, the method of Patent Document 4 does not describe a method for evaluating the positional deviation of the aggregate channel box from the normal position due to the deformation of the channel box by neutron irradiation. There was a problem that the thermal deformation could not be calculated by predicting box deformation.

  In addition, if combustion proceeds while the assembly is misaligned due to channel box deformation, the combustion history effect and burnup distribution effect of changes in the neutron spectrum in the assembly due to changes in the water gap width accumulate, and the assembly is instantaneously misaligned. A difference from the effect of doing so occurs. This history effect generally occurs in a direction that cancels out the instantaneous effect. However, in Patent Document 4, this history effect is not considered at all.

  An object of the present invention is to provide a core performance calculation apparatus capable of accurately calculating core nuclear characteristics in consideration of instantaneous and hysteresis effects when offset bundles are adjacent in a D lattice core or when a channel box is deformed. .

According to the first aspect of the present invention, in the core performance calculation of the boiling water reactor in which the D lattice core and the offset bundle are loaded on the D lattice core, the critical eigenvalue of the core, neutron Calculates the neutron flux and neutron flow at the node interface where each assembly is homogenized based on the diffusion power calculation means for calculating the bundle distribution and power distribution, and the diffusion theory using the neutron flux discontinuity factor Node boundary value calculation means, assembly cross-sectional average power calculated by the in-core power distribution calculation means, and sensitivity coefficient of fuel rod local output to neutron flux and neutron flow at each surface of the aggregate boundary by the node boundary value calculation means And a linear power density calculating means for calculating the linear power density of the assembly based on the local power of the infinite lattice fuel rod corrected by using.

  According to the first 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 diffusion model using the nodal average nuclear constant. The node boundary value calculation means calculates the neutron flux and neutron flow at the node boundary surface obtained by homogenizing each aggregate based on the diffusion theoretical calculation using the neutron flux discontinuity factor, and the line power density calculation means Infinite lattice fuel corrected using the average power of the cross section of the aggregate calculated by the means for calculating the power distribution in the reactor and the sensitivity coefficient of the local power of the rod to the neutron flux and neutron flow on each side of the aggregate boundary by the means for calculating the node boundary value Calculate the line power density of the aggregate from the local output of the bar.

The invention of claim 2 is based on a diffusion model using a node average nuclear constant in a core performance calculation of a boiling water reactor in which a D lattice bundle and an offset bundle are loaded on a D lattice core. Calculates the neutron flux and neutron flow at the node interface where each assembly is homogenized based on the diffusion power calculation means for calculating the bundle distribution and power distribution, and the diffusion theory using the neutron flux discontinuity factor Node boundary value calculation means, assembly cross-sectional average power calculated by the in-core power distribution calculation means, and sensitivity coefficient of local fuel rod output to neutron flux and neutron flow at each surface of the aggregate boundary by the node boundary value calculation means Calculated by the linear power density calculating means for calculating the linear power density of the assembly by the infinite lattice fuel rod local power corrected using And a critical power ratio calculating means for calculating a critical power ratio of the fuel assembly from the fuel rod R factor calculated using the average fuel cross section output and the fuel rod local power by the linear power density calculating means. is there.

  In the invention of claim 2, the in-core power distribution calculating means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the diffusion model using the nodal average nuclear constant. And the line power density calculating means, based on the diffusion theory calculation using the neutron flux discontinuity factor, a node boundary value calculating means for calculating the neutron flux and neutron current of the node boundary surface homogenized each aggregate, Infinite Lattice Fuel Rod Corrected by Sensitivity Factor of Local Power of Fuel Rod to Neutron Flux and Neutron Flow on Each Face of Aggregate Boundary by Means of Node Boundary Value Calculation Means by Aggregate Cross Section Average Power Calculated by Reactor Power Distribution Calculation Means The line power density of the aggregate is calculated from the local output. Further, the limit power ratio calculation means is configured to calculate the fuel assembly from the fuel rod R factor calculated by using the assembly cross-sectional average output calculated by the in-core power distribution calculation means and the fuel rod local output by the linear power density calculation means. Calculate the limit power ratio.

According to the invention of claim 3, in the core performance calculation of the boiling water reactor in which the D lattice core and the offset bundle are loaded on the D lattice core, the critical eigenvalue of the core, neutron Calculates the neutron flux and neutron flow at the node interface where each assembly is homogenized based on the diffusion power calculation means for calculating the bundle distribution and power distribution, and the diffusion theory using the neutron flux discontinuity factor Sensitivity coefficient of fuel rod R factor to node boundary value calculation means, aggregate cross-sectional average power calculated by said reactor power distribution calculation means, and neutron flux and neutron flow at each surface of the aggregate boundary by said node boundary value calculation means And a limit power ratio calculating means for calculating the limit power ratio of the aggregate using the infinite lattice R factor corrected by using.

In the invention of claim 3, the reactor power distribution calculation means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the diffusion model using the node average nuclear constant, and the node boundary value calculation means The neutron flux and neutron flow at the node interface where each aggregate is homogenized are calculated based on the diffusion theory using the bundle discontinuity factor. The critical power ratio calculating means is a sensitivity coefficient of the fuel rod R factor to the neutron flux and neutron flow at each surface of the aggregate boundary by the aggregate cross-sectional average power calculated by the in-core power distribution calculating means and the node boundary value calculating means. The limit output ratio of the aggregate is calculated by the infinite lattice R factor corrected by using.

According to the invention of claim 4, in the core performance calculation of the boiling water reactor in which the D lattice core and the offset bundle are loaded on the D lattice core, the critical eigenvalue of the core, the neutron, based on the diffusion model using the nodal average nuclear constant Calculates the neutron flux and neutron flow at the node interface where each assembly is homogenized based on the diffusion power calculation means for calculating the bundle distribution and power distribution, and the diffusion theory using the neutron flux discontinuity factor In-reactor position neutron flux with respect to node boundary value calculation means, aggregate cross-sectional average power calculated by the in-core power distribution calculation means, and neutron flux and neutron flux at each surface of the aggregate boundary by the node boundary value calculation means In-core neutron flux counter count and in-core neutron flux counter count value calculation from infinite lattice gauge position neutron flux corrected using sensitivity coefficient And it has a door.

  In the invention of claim 4, the power distribution calculation means in the reactor calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the diffusion model using the node average nuclear constant, and the node boundary value calculation means The neutron flux and neutron flow at the node interface where each aggregate is homogenized are calculated based on the diffusion theory using the bundle discontinuity factor. Then, the in-core neutron flux measuring device count value calculation means includes an in-reactor average power output by the in-reactor power distribution calculation means and a neutron flux and neutron flow at each surface of the aggregate boundary by the nodal boundary value calculation means. The in-core neutron flux counter count and irradiation dose are calculated from the infinite lattice instrument position neutron flux corrected using the sensitivity coefficient of the instrument position neutron flux.

According to the invention of claim 5, in the core performance calculation of the boiling water reactor in which the D lattice core and the offset bundle are loaded on the D lattice core, the critical eigenvalue of the core, neutron Calculates the neutron flux and neutron flow at the node interface where each assembly is homogenized based on the diffusion power calculation means for calculating the bundle distribution and power distribution, and the diffusion theory using the neutron flux discontinuity factor Sensitivity of control rod position neutron flux to node boundary value calculation means, aggregate cross-sectional average power calculated by the reactor power distribution calculation means, and neutron flux and neutron flow at each surface of the aggregate boundary by the node boundary value calculation means A control rod irradiation amount calculating means for calculating the control rod irradiation amount from the infinite lattice control rod position neutron flux corrected using the coefficient.

  In the invention of claim 5, the in-core power distribution calculation means calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the diffusion model using the nodal average nuclear constant, and the nodal boundary value calculation means The neutron flux and neutron flow at the node interface where each aggregate is homogenized are calculated based on the diffusion theory using the bundle discontinuity factor. Then, the control rod irradiation amount calculation means calculates the cross-sectional average power of the aggregate calculated by the in-core power distribution calculation means and the neutron flux on each surface of the aggregate boundary by the node boundary value calculation means and the control rod position neutron flux for the neutron flow. The control rod dose is calculated from the infinite lattice control rod position neutron flux corrected using the sensitivity coefficient.

  According to a sixth aspect of the present invention, in the first to fifth aspects of the invention, the in-core power distribution calculating means calculates a fast neutron flux distribution by a modified first group diffusion calculation, and the obtained fast neutron flux and infinite lattice The thermal neutron flux and the thermal neutron flow are calculated using the spectral index, which is the ratio of the thermal neutron flux to the fast neutron flux in the calculation.

  In the invention of claim 6, in addition to the effects of the inventions of claims 1 to 5, the fast neutron flux distribution is calculated by the modified first group diffusion calculation, and the obtained fast neutron flux and the thermal neutron flux in the infinite lattice calculation are calculated. Calculate the thermal neutron flux and thermal neutron flow using the spectral index, which is the ratio to the fast neutron flux.

  According to a seventh aspect of the present invention, in the first to sixth aspects of the present invention, the calculated value of the in-core neutron flux count value obtained by correcting the adjacent effect between the D lattice bundle and the offset bundle is used as the in-core neutron flux count In-furnace power distribution correction means for correcting the in-furnace power distribution calculation value is provided by adapting to the actual measurement value.

  In the invention of claim 7, in addition to the effects of the inventions of claims 1 to 6, the calculated value of the in-core neutron flux measuring instrument value corrected for the adjacent effect of the D lattice bundle and the offset bundle is calculated as the in-core neutron The power distribution distribution calculation value in the furnace is corrected according to the measured value of the bundle measuring instrument.

The invention of claim 8 is the fuel of claim 1 to claim 7 , wherein the fuel is obtained based on the burnup distribution in the fuel assembly and the neutron flux and neutron flow at each surface of the assembly boundary by the node boundary value calculation means. Using the spectral history distribution, which is the burnup average of the ratio of thermal neutron flux to fast neutron flux at each fuel rod position in the assembly, the node average nuclear determination due to the hysteresis effect associated with the adjacency between the D lattice bundle and the offset bundle. The correction amount of the number, the fuel rod local power distribution, and the fuel rod R-factor distribution is calculated.

In the invention of claim 8, in addition to the effects of the inventions of claims 1 to 7, the burnup distribution in the fuel assembly and the spectral history distribution which is the average burnup value of the ratio of thermal neutron flux to fast neutron flux are used. Then, the correction amount of the node average 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 core performance calculation apparatus of the present invention, the calculation of the core characteristics considering the effect of the adjacent D-lattice bundle and the offset bundle is the multi-assembly detailed calculation corresponding to the combination of the D-lattice bundle and the offset bundle. The calculation is based on the correction table prepared in advance, the diffusion node method using aggregate discontinuity factors, or the analytical diffusion model, so that the critical eigenvalue, reactor power distribution, maximum linear power density, and critical power ratio of the core It has the excellent effect of accurately calculating the in-core neutron detector count, the detector irradiation amount, the control rod irradiation amount, and the like.

  In addition, the channel box deformation caused by fast neutron irradiation can be estimated by integrating the fast neutron dose on each surface of the channel box by core power distribution calculation and predicting the channel box deformation caused by neutron irradiation based on the irradiation growth model. Therefore, it is possible to calculate the core characteristics considering the channel box deformation. In addition, the calculation of the core characteristics of the core considering channel box deformation is based on the channel table deformation amount of the aggregate of interest and the adjacent aggregate based on the correction table or diffusion model prepared in advance by multi-assembly detailed calculation. Since the hysteresis effect is also corrected, the core critical eigenvalue, reactor power distribution, maximum linear power density, critical power ratio, reactor neutron detector count and detector dose, control rod dose, etc. are accurately calculated. There is an excellent effect that can be done.

  The best mode for carrying out the present invention will be described below with reference to the accompanying drawings.

  Embodiments of the present invention will be described below. First, in the D-lattice core, core nuclear characteristics such as aggregate average nuclear constant, fuel rod local power, fuel rod R factor, in-core neutron detector count, and control rod irradiation amount are calculated with good accuracy considering the offset bundle adjacent effect. The case where it does is demonstrated. That is, this is achieved by the following processes [1], [2], and [3].

[1] Changes in aggregate nucleus characteristics due to the adjoining of the D lattice bundle and the offset bundle are calculated based on the multi-assembly detailed calculation model or the neutron diffusion model as shown in the following (a) to (d). .

(A) Multi-aggregate detailed two-dimensional calculation model.
Aggregate average nuclear constant, fuel rod output distribution, fuel rod R-factor distribution, etc. due to adjacent D lattice bundle and offset bundle in advance by multi-assembly detailed two-dimensional calculation for the target assembly and adjacent assembly The core characteristic change amount is directly prepared as a table of combinations of aggregates having different water gap widths adjacent to the target aggregate.

(B) Analytical diffusion model.
Based on the analytical solution of the two-region diffusion model in which the water gap and the fuel region are homogenized, the change in the thermal neutron flux distribution in the assembly is obtained. Evaluate the amount of change.

(C) Single assembly neutron flux discontinuity factor diffusion model.
Based on the homogeneous diffusion nodal method using the aggregate neutron flux discontinuity factor, the homogeneous neutron flux distribution in the node that homogenizes each aggregate is obtained. 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 inhomogeneous 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 the "inhomogeneous heat in the node" Calculate the inhomogeneous thermal neutron flux distribution in the node by multiplying the factor of the ratio of the neutron flux to the inhomogeneous thermal neutron flux as the correction factor for the transient component due to thermal neutron diffusion of the homogeneous thermal neutron flux distribution. The aggregate nuclear properties are calculated from this inhomogeneous thermal neutron flux distribution. As the homogeneous diffusion model, a multigroup node method, a modified first group node method, or the like is used.

(D) Single assembly neutron flux discontinuity factor boundary perturbation model.
Based on the homogeneous diffusion nodal method using the aggregate neutron flux discontinuity factor, the neutron flux and neutron current at the node boundary obtained by homogenizing each aggregate are obtained.

  Using the change in aggregate nuclear characteristics for changes in neutron flux and neutron flow at the aggregate boundary evaluated in the infinite lattice calculation as the sensitivity coefficient, the aggregate nuclear characteristics from the neutron flux and neutron current at the above node boundary are used. Calculate the change.

  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 limit power ratio calculation, in addition to directly calculating the R factor change in the assembly, 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 core characteristics considering the offset bundle adjacency are calculated based on the table and diffusion model prepared by the detailed multi-assembly calculation, and the combustion history effect is also considered. Changes in nuclear properties can be calculated accurately. In addition, the critical eigenvalue of the core, the power distribution in the reactor, the count value of the reactor neutron detector, the detector irradiation amount, the control rod irradiation amount, etc., which have been ignored in the past, can be taken into consideration.

  FIG. 1 is a block diagram showing a first embodiment of the present invention. The aggregate average nuclear constant calculation means 11 uses the correction amount for the aggregate average nuclear constant or the infinite lattice average nuclear constant evaluated in advance by multi-aggregate detailed calculation according to the combination of the adjacent D lattice bundle and offset bundle. The average nuclear constant is calculated. That is, the node average nuclear constant is calculated as a function of the burnup ratio, void ratio, etc. of the node, and at that time, the aggregate average nuclear constant evaluated in advance by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and offset bundle. Use the number to take into account the effects of neighboring aggregates.

  The in-core power distribution calculating means 12 calculates the critical eigenvalue, neutron flux distribution, and power distribution of the core based on the core diffusion calculation using the nuclear constant calculated by the aggregate average nuclear constant calculating means 11. The average nuclear constant is used to calculate the neutron flux distribution and core eigenvalue in the reactor 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 described later, and the like can be used. The power distribution in the reactor is calculated based on the neutron flux distribution, and the aggregate cross-section average power is calculated.

  The local output distribution calculating means 13 calculates a correction amount for the fuel rod local output in the fuel assembly or the infinite lattice fuel rod local output evaluated in advance by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and the offset bundle. Used to calculate the fuel rod local power.

  Next, the line power density calculating unit 14 uses the assembly cross-sectional average output calculated by the in-core power distribution calculating unit 12 and the fuel rod local output calculated by the local power distribution calculating unit 13 to output the line output of the assembly. Calculate the density. In this linear power density calculation means 14, as a thermal limit value of the core, a 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. The maximum linear power density of each fuel assembly cross section is calculated using the maximum local power peaking coefficient of the body and the average power of the cross section of the assembly based on the power distribution calculation in the core. In this case, the accuracy can be further improved by using the output obtained by correcting the offset bundle adjacency effect as the aggregate section average output by the in-core power distribution calculation means 12.

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

  Next, the in-reactor neutron flux measuring device count value calculation means 16 determines the thermal neutron flux at the in-reactor neutron measuring device position 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 in-reactor neutron detector count and the detector dose. Then, the in-core power distribution correction means 17 adapts the calculated value of the in-core neutron flux measuring device count value corrected for the adjacent effect of the D lattice bundle and the offset bundle to the actually measured value of the in-core neutron flux measuring device. To correct the power distribution calculation value in the furnace. The in-core power distribution calculated value can be corrected 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.

  Further, the control rod irradiation amount calculation means 18 calculates the control rod irradiation amount using the control rod position thermal neutron flux evaluated in advance by the multi-assembly detailed calculation according to the combination of the adjacent D lattice bundle and the offset bundle. .

  Next, multi-aggregate calculation is performed to calculate the average nuclear constant, the discontinuity factor, the fuel rod local power, the fuel rod R factor, the thermal neutron flux at the neutron measuring instrument, etc. The method 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 determined by a combination of the adjacent D lattice bundle and the offset bundle. FIG. 2 shows eight possible patterns for combinations of adjacent bundles centered on the offset bundle. In addition, since the influence of the bundle adjacent from the diagonal diagonal direction with respect to the central bundle of interest 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. For each such pattern, multi-aggregate calculation may be performed to evaluate the aggregate nucleus characteristics.

Proceeding combustion remains aggregates misaligned, combustion history effect and combustion distribution effects due to neutron spectrum variation in the aggregation by water gap width variation is accumulated, when momentarily aggregates misaligned A difference from the effect occurs. In order to incorporate this history effect, it is necessary to perform combustion calculation in multi-aggregate calculation.

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

  Although the number of combination patterns of the D lattice and the offset bundle is not so large, in actuality, the number of combinations increases and the calculation cost increases when considering the difference in burnup and void ratio of each bundle. However, the combination of burnups can be approximated by a combination of burnups of typical batch types if the offset bundle is introduced step by step for each cycle.

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

  Furthermore, as a method for simplifying the correction calculation, a fuel that has been evaluated in advance by a multi-assembly detailed calculation for a combination of a D lattice bundle and an offset bundle that has the strictest thermal limit values such as the maximum linear power density and the limit power ratio. Using the correction amount for the infinite lattice calculation value of the fuel rod local power and fuel rod R factor in the assembly for all D lattice bundles and offset bundles, the core power density and the critical power ratio of the core are corrected by core diffusion calculation. You can also

(B) Analytical diffusion model.
Next, there is provided a thermal neutron flux change calculating means for calculating the change of the thermal neutron flux distribution in the aggregate with respect to the amount of change of the water gap width of the aggregate of interest from the infinite lattice system, based on the analytical model for the diffusion equation. Changes in the thermal neutron flux distribution in the aggregate due to substantial changes in the water gap width of the aggregate due to the adjacent lattice bundle and offset bundle are obtained based on an analytical model for the diffusion equation. Then, the nuclear property change amount is evaluated from the thermal neutron flux distribution.

  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, an 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 following expression (1).

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

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

(C) Single assembly neutron flux discontinuity factor diffusion model.
Next, an inhomogeneous thermal neutron flux calculating means for calculating the inhomogeneous thermal neutron flux distribution in the aggregate is provided, and the effect obtained by the difference in water gap width between the D lattice bundle and the offset bundle is obtained by the infinite lattice calculation. Consider using neutron flux discontinuity factor. Therefore, multi-aggregate calculation is not required. In this method, the neutron flux distribution in the aggregate is obtained by the diffusion node method using the neutron flux discontinuity factor for the system consisting of nodes that homogenize each aggregate, and the amount of change in the nuclear characteristics of the aggregate is calculated from this neutron flux distribution. To evaluate.

  That is, based on the diffusion theory using neutron flux discontinuity factor, we calculated the homogeneous thermal neutron flux distribution in the node where each aggregate was homogenized, and evaluated the inhomogeneous heat at the boundary of the aggregate evaluated in the infinite lattice calculation. Correction of transient components due to thermal neutron diffusion of homogeneous thermal neutron flux distribution by multiplying the factor of "ratio of neutron flux to homogeneous thermal neutron flux" and "ratio of homogeneous thermal neutron flux to non-homogeneous thermal neutron flux in the assembly" It is a coefficient. This calculates the inhomogeneous thermal neutron flux distribution in the assembly. Then, the core eigenvalue, the in-core neutron flux distribution, and the power distribution are calculated by correcting the assembly average nuclear constant using the inhomogeneous neutron flux distribution by the inhomogeneous thermal neutron flux calculation means.

  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. The multi-group diffusion node method itself using discontinuous factors is already known. For example, in the document "Assembly Homogenization Techniques for Light Water Reactor Analysis," KSSmith, Progress in Nuclear Energy, vol. 17, p303, 1986. Are listed.

  An example of a method for calculating the local power distribution in the assembly using the neutron flux distribution obtained by the node method calculation is "SIMULATE-3 Pin Power Reconstruction Methodology and Benchmarking," KRRempe et.al, Proceedings of lnternational Reactor Physics Conference, 111-19, Jackson Hole, 1988.

  The D lattice bundle and the offset bundle have different water gap widths, so the neutron flux distribution in the aggregate is different, and the aggregate neutron discontinuity factor is also different. 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 the nodes where the aggregates are homogenized in the diffusion node method. That is, for example, the following boundary conditions for the neutron flux are given at the x-direction boundary between the target assembly n and the adjacent assembly m.

  Here, f represents a neutron flux discontinuity factor at the boundary surface, and Ψ represents a neutron flux at the boundary surface. If the aggregate of interest and the adjacent aggregate are the same lattice type in an infinite lattice system, the neutron discontinuity factors are also equal, and the above boundary condition merely represents 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 is generated at the node boundary based on the above boundary condition. The change in neutron flux distribution in the aggregate due to the difference can be calculated. In the multi-group diffusion node method, the nodal average thermal neutron flux and the nodal boundary thermal neutron flux are given by the whole core calculation using the neutron flux discontinuity factor for the thermal neutron flux.

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

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 Ψ obtained by the 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.

  Here, the symbol ∞ represents an infinite lattice calculation value.

  However, when the inventors applied this method to the offset bundle system, the nodal average neutron flux, the nodal boundary neutron flux, and the nodal boundary neutron flow, the results of the diffusion node method agree well with the multi-aggregate detailed calculation. Nevertheless, it was found that there was a large error in the local fuel rod output. This is due to the assumption 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 infinite lattice calculation. It was found that this was not true.

  In order to accurately calculate the inhomogeneous thermal neutron flux distribution from the in-node homogeneous thermal neutron flux, the inventors have evaluated "the inhomogeneous thermal neutron flux at the aggregate boundary to the homogeneous thermal neutron flux in the infinite lattice system. Ratio "and" ratio of homogeneous thermal neutron flux to inhomogeneous thermal neutron flux in the aggregate "are used as correction factors for transient components due to thermal neutron diffusion in the homogeneous thermal neutron flux distribution. We found that the inhomogeneous thermal neutron flux distribution in the node can be calculated with high accuracy.

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

  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 using the load average for the four aggregate boundary surfaces.

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

  The fuel rod R-factor distribution can be calculated from the fuel rod local power distribution, and the limit power ratio can be calculated from this.

  The aggregate neutron flux discontinuity factor on each surface of the aggregate is prepared for each aggregate in advance by 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 assembly discontinuity factor with respect to the thermal neutron flux.

  As an example showing the calculation accuracy of this method, as in the conventional FIG. 19 showing the effect of adjacent offset bundles, the assembly in the case where two offset bundles are adjacent to each other with a control rod sandwiched around the D lattice bundle. The local output distribution is calculated by this method, and the result compared with the reference solution by the multi-aggregate detailed calculation is shown in FIG. 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 maximum local peaking coefficient of the aggregate changes by about 10% due to the adjacency of the offset bundle, but this method reproduces this well.

  This method does not require detailed multi-assembly calculations when preparing a table of assembly neutron flux discontinuity factors, and the neutron flux discontinuity factors for each surface of the fuel assembly are used as a function of each surface of the target assembly. It may be calculated independently. As described above, this method has an advantage that the preparation and reference of the table are simpler than the above-described method that requires 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 from the core calculation and the “ratio of thermal neutron flux to fast neutron flux (spectral index)” in the infinite lattice system. When the spectral index is different between adjacent assemblies, thermal neutron spatial movement occurs, but the change from the infinite lattice of the thermal neutron flux is caused by a system consisting of nodes in which the target assembly and adjacent assemblies are homogenized. On the other hand, it is calculated by applying a diffusion model.

  In the modified first group node method, the homogeneous thermal neutron flux in the node is obtained by analytically solving the diffusion equation of this system. Details of this method are described in, for example, the document “Verification of LOGOS Nodal Method with Heterogeneous Burnup Calculations for a BWRcore,” T. Iwamoto et al., Transaction of American Nuclear Society, vol. 71, p251, 1994.

  Even in the modified first group method, the thermal neutron flux change due to channel box deformation uses a neutron flux discontinuity factor as a boundary condition for the thermal neutron flux in a system consisting of nodes of homogenized aggregates of interest and adjacent aggregates. Thus, the calculation can be performed based on the same principle as in the above-described multigroup 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 due to the difference in water gap width by providing node boundary value calculation means to calculate the neutron flux and neutron flow at the node boundary surface where each assembly is homogenized based on diffusion theory calculation using neutron flux discontinuity factor. Using the difference of the aggregate discontinuity factor, the neutron flux and neutron flow at the boundary of the aggregate are obtained, and the change of the boundary value from the infinite lattice and the sensitivity coefficient of the characteristic of the aggregate nucleus to the aggregate nucleus are used. Calculate the change in properties. As described above, this is because the diffusion node method using the neutron flux discontinuity factor pays attention to high accuracy for the neutron flux and neutron flow at the aggregate boundary.

  Based on the boundary perturbation method, an example of calculating local peaking in an aggregate using “neutron flow / neutron flux” as the perturbation at the aggregate boundary is shown in the literature “A Boundary Condition Perturbation Method for Prediction of Pin Power 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) within the assembly for the fuel rod (x, y) is calculated by the following equation (8).

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

  However, in the Rahnema literature, when the neutron current / neutron flux at the interface is calculated based on the diffusion theory, the aggregate neutron flux discontinuity factor is not used, The change in neutron flow / neutron flux due to local neutron flux distribution change near the water gap cannot be calculated.

Therefore, the node boundary value Γ is a multi-group diffusion node method using a neutron flux discontinuity factor for a system consisting of nodes with homogenized aggregates of interest and adjacent aggregates as in the diffusion node method described above. Apply to find. Alternatively, in the modified first group difference method, the nodal boundary thermal neutron flux may be calculated from the spectral index of adjacent nodes 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).

  Equation (8) shows the correction calculation method by the boundary perturbation method for the local output of the assembly. However, the correction calculation of other assembly nuclear characteristic quantities is also a sensitivity coefficient for the neutron flow / neutron flux of each nuclear characteristic quantity. Can be calculated in a similar manner.

  If combustion proceeds while the fuel assembly is misaligned, the history effect of changes in the neutron spectrum in the assembly due to the change in the water gap width and the effect of the burnup distribution are accumulated, and the effect of instantaneous misalignment of the assembly The difference occurs. This is called the spectral history effect because the neutron spectrum (ratio of thermal neutron flux to fast neutron flux) changes from the infinite system, resulting in an increase in uranium combustion delay and plutonium accumulation.

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

  In the present invention, this technique is applied to the spectral history effect due to the adjacent offset bundle, and the burnup distribution and the spectral history distribution in the fuel assembly corrected for the effect of changing the thermal neutron flux distribution due to the change in the water gap width are calculated and the combustion is performed. Calculate the correction amount for the average nuclear constant of the assembly, local distribution of fuel rods, etc. due to the hysteresis effect. For example, for the fuel rod local output, the spectrum history SH (x, y) at the fuel rod position (x, y) in the assembly is expressed by the following equation (10).

Where E is the burnup, Ψ ₂ , Ψ ₁ are the thermal neutron flux and fast neutron flux at the position (x, y) calculated considering the offset bundle adjacency, and ∞ is the infinite lattice system Indicates the value of. The correction of the local output distribution LPF considering the spectrum history is given by the following equation (11) by using the spectrum history correction coefficient (係数 Σ / ∂SH) for the thermal group fission cross section Σ.

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

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

[1] The fast neutron dose or burnup on each face of the channel box is integrated by core performance calculation, and the irradiation growth on the opposite face of the channel box is calculated as a function of the fast neutron dose or burnup on each face. The amount of deformation of the channel box is calculated from the difference in the amount of irradiation growth on the opposing surfaces.
[2] Aggregate nuclear properties such as aggregate average nuclear constant, local fuel rod output, neutron detector position thermal neutron flux, control rod position thermal neutron flux, taking into account channel box deformation in the target aggregate and adjacent aggregates Calculate
[3] In order to calculate the change in the above-described aggregate nucleus characteristics from the deformation amount of the channel box, it is based on the following multi-assembly detailed calculation model or neutron diffusion model.

(A) Multi-aggregate detailed two-dimensional calculation model.
By performing multi-assembly detailed two-dimensional calculation for the target aggregate and adjacent aggregates, it is possible to directly determine the amount of change in core characteristics such as the average average nuclear constant, fuel rod output distribution, fuel rod R factor distribution, etc. Prepare as a table of deformation amount of water gap width on the surface. Here, multi-aggregate detailed two-dimensional calculation is a detailed calculation that considers the inhomogeneity of each aggregate as it is for a system that simulates a part of the core by combining a small number of aggregates (usually 4 to 16). This is a method for evaluating in detail the influence of adjacent aggregates on the aggregate of interest.

(B) Single assembly neutron flux discontinuity factor diffusion model.
The amount of change in the aggregate neutron flux discontinuity factor is calculated from the channel box deformation, and the thermal neutron flux distribution in the aggregate is calculated based on the homogeneous diffusion model for the nodes where each aggregate is homogenized. The nuclear property change is evaluated from the bundle 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 of the assembly neutron flux discontinuity factor is prepared in advance by a single assembly detailed calculation as a channel box deformation amount table. As the homogeneous diffusion model, a multigroup node method, a modified first group node method, or the like can be used.

(C) Single assembly neutron flux discontinuity factor boundary perturbation model.
Based on the homogeneous diffusion nodal method using the aggregate neutron flux discontinuity factor, the neutron flux and neutron current at the node boundary obtained by homogenizing each aggregate are obtained.

  The change in aggregate nuclear properties from the neutron flux and neutron current at the above node boundary is used as the sensitivity coefficient, which is evaluated in the infinite lattice system, using the change in aggregate nuclear properties for the change in neutron flux and neutron current at the aggregate boundary. Calculate

  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.
Based on the analytical solution of the two-region diffusion model in which the water gap and fuel region are homogenized from the channel box deformation, the change in the thermal neutron flux distribution in the assembly is obtained, and the nuclear property change is evaluated from this thermal neutron flux distribution. .

[4] In the limit power ratio calculation considering channel box deformation, in addition to directly determining the R factor change in the assembly, it can also be calculated from the fuel rod output change.
[5] Combustion history effect associated with channel box deformation is corrected using a spectral history distribution defined as the burnup distribution within the assembly accompanying the channel box deformation and the burnup average value of the ratio of thermal neutron flux to fast neutron flux To do.

  As described above, in the second embodiment, the fast neutron irradiation amount or the burnup degree on each surface of the channel box is integrated by the core performance calculation. Therefore, the channel box deformation amount due to the neutron irradiation is predicted and evaluated. It becomes possible to calculate the core characteristics considering the deformation.

  In addition, the calculation of core characteristics taking into account channel box deformation is calculated based on the table and diffusion model prepared by the detailed multi-assembly calculation from the channel box deformation of the target assembly and the adjacent assembly. Therefore, the change in nuclear characteristics can be calculated with high accuracy.

  In addition, the influence of channel box deformation on the critical eigenvalue of the core, the power distribution in the reactor, the count value of the neutron detector in the reactor, the detector irradiation amount, the control rod irradiation amount, etc., which have been ignored in the past can be taken into consideration.

  FIG. 6 is a block diagram showing the second embodiment of the present invention. The in-core power distribution calculation means 19 inputs core state data and calculates the neutron flux distribution and core eigenvalue in the core 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, or the like is used. The power distribution in the reactor is calculated based on the neutron flux distribution, and the aggregate cross-section average power is calculated.

The channel box dose calculation means 20 calculates the fast neutron dose on each surface of the channel box of the aggregate based on this neutron flux distribution. Here, the fast neutron flux [psi _n is in the plane n, the multigroup diffusion node method is calculated revealed, also in the modification 1 group difference model using polynomial node average neutron flux focused aggregate and surrounding assembly and out It can be calculated by inserting. Accordingly, the fast neutron irradiation amount φ can be calculated as shown in the equation (12) by time integration of the fast neutron flux. Here, dt is a time width.

  The channel box deformation amount calculation means 21 evaluates the deformation amount of the channel box by neutron irradiation using the irradiation growth model using the fast neutron irradiation amount of each surface. As an evaluation method, it can be calculated from the irradiation growth amount in the axial direction on the opposing surfaces of the channel box.

In the channel box deformation amount calculation method, the channel box deformation amount is analytically calculated by assuming that the curvature of the channel box in the axial direction has a uniform curvature in accordance with the difference in irradiation growth amount between the two opposing surfaces A and B. That is, as shown in FIG. 7, the average fast neutron doses in the axial direction of the surface A and B of the channel box are φ _A and φ _B , respectively, and the curvature of the bending in the axis (Z) direction of the channel box is the bending. Is expressed by the following equation (13). Then, the amount of bending at the height Z in the axial direction is expressed as in equation (14) by integrating equation (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 by the equations (16) and (17). An example of this function form is shown in FIG.

  Here, a, b, c, d, and e are fitting coefficients. This coefficient can be obtained from an empirical formula. This coefficient is determined by using the measured values of channel box deformation in the past in order to correct errors due to random bending due to release of residual stress during channel manufacturing and variations in irradiation growth of Zircaloy. Also good.

As a calculation method of the channel box deformation amount, there is 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 the straight line connecting the upper and lower ends of the A surface or B surface of the channel and each node can be obtained as the bending amount δ ⁱ .

  Note that fast neutrons that cause irradiation growth are neutrons having an energy of 1 MeV or more, and do not necessarily match all fast neutrons given in the core performance calculation. Therefore, the ratio f of the fast neutrons used in the fast neutron dose calculation to the total fast neutrons can be calculated as a table of burnup E and void ratio V as shown in the following equation (18).

  In the second embodiment, the core performance calculation apparatus is configured using an irradiation growth model that uses a one-to-one correspondence between the fast neutron dose and the burnup. In this case, instead of the channel box dose calculation means 20 of FIG. 1, burnup distribution calculation means is provided.

  Next, a method for calculating the burnup on each face of the aggregate channel box in the burnup distribution calculation means will be described. Here, the burn-up En in the plane n is calculated by time-integrating the power density at the channel box position because the intra-node power distribution is calculated with the multi-group diffusion node method. Further, in the modified first group difference model, the node average burnup of the aggregate of interest and the surrounding aggregate can be calculated by interpolation using a polynomial.

  In the model using the burnup, the irradiation growth distortion amount of the channel box due to the neutron irradiation can be calculated as a function of the burnup E of each surface, for example, as in the following equations (19) and (20).

  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 measurement value of the channel box deformation amount in the past in order to correct an error depending on uncertainty at the time of manufacturing the channel. FIG. 10 shows an example of the function form of ε (E).

  Next, the in-reactor power distribution recalculation means 22 uses the channel box deformation amount evaluated by the channel box deformation amount calculation means 21 as described above, and the assembly nuclear constant and the neutron flux with the channel box deformation effect corrected. The discontinuity factor is calculated and used to recalculate the critical eigenvalue of the core, the power distribution in the reactor, and the fuel assembly cross-sectional average power distribution.

  In the cotton power density calculation means 23, as the thermal limit value of the core, 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 average cross-section of the aggregate by calculating the power distribution in the core Using the output, the maximum linear power density of each fuel assembly cross section is calculated. Instead of obtaining the maximum local output peaking coefficient of the assembly from the fuel rod local output distribution corrected for the channel box deformation effect, the accuracy is reduced, but the channel directly corresponds to the maximum local output peaking coefficient of the assembly. The box deformation effect may be corrected.

  Further, the limit power ratio calculation means 24 uses the aggregate maximum R factor obtained from the fuel rod R factor distribution considering the channel box deformation and the aggregate average output to burn out the fuel rods of each fuel assembly. The marginal power ratio, which is a monitoring index, is calculated. 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 assembly 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 assembly. It may be corrected.

  In calculating these linear power density, critical power ratio, etc., the fuel assembly average output recalculated using the nuclear constant corrected for the channel box deformation is used as the fuel assembly average output. Accuracy can be improved.

Next, the in-core neutron flux measuring device count value calculation means 25 calculates the in-core neutron detector count value and the irradiation amount using the thermal neutron flux at the in-core neutron measuring device position considering the channel box deformation. .
An example of a counting calculation method in the case of a movable in-core neutron flux detector (TIP) as an in-core neutron flux detector is described in the aforementioned Wooly document. The detector dose is given by the time integration of the thermal neutron flux at the position of the in-core neutron measuring instrument.

  Although not shown in FIG. 6, the in-core output is obtained by learning the error of the calculated value of the in-core neutron detector count value corrected for channel deformation based on the method of the Tuiki literature. It is also possible to correct the distribution calculation value. Further, the control rod irradiation amount calculation means 26 calculates the control rod irradiation amount by time integration of the control rod position thermal neutron flux corrected for the channel box deformation effect.

  Next, details on three different methods to calculate the average nuclear constant, the discontinuity factor, the fuel rod local power, the fuel rod R factor, the thermal neutron flux at the neutron measuring instrument, etc. Explained.

(A) Multi-aggregate detailed two-dimensional calculation model.
First, as method 1, a table of nuclear characteristic changes such as the average nuclear constant of the aggregate, the local output of the fuel rod, the fuel rod R factor, the thermal neutron flux at the position of the neutron measuring instrument with respect to the channel box deformation of the aggregate of interest and the adjacent aggregate Or it prepares as a fitting type | formula and calculates the nuclear characteristic change in the fuel assembly which correct | amended the channel box deformation effect directly. A table or fitting equation for the change in the characteristic amount of the target aggregate with respect to the channel box deformation amount of the target aggregate and the adjacent aggregate is evaluated by multi-aggregate two-dimensional detailed calculation.

  Here, multi-aggregate detailed two-dimensional calculation is a detailed calculation that considers the inhomogeneity of each aggregate as it is for a system that simulates a part of the core by combining a small number of aggregates (usually 4 to 16). This is a method for evaluating in detail the influence of adjacent aggregates on the aggregate of interest.

  This table or fitting equation is a function of not only the deformation amount of each surface of the aggregate of interest but also the deformation amount of the adjacent surface of the adjacent aggregate, and is generally a complex function for multivariate combinations. On the other hand, Patent Document 4 describes that the calculation is merely based on the amount of misalignment of the fuel assembly, and the evaluation method is not clear at all. In this method, this function is organized as a table of water gap widths for each face of the channel box of the aggregate of interest. This table is also a multivariate function (tetravariate function) with respect to 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 deformation of the channel box of the aggregate adjacent to the aggregate of interest and only the channel box 2b of the aggregate of interest is deformed. As shown in FIG. 11, the water gap deformation amount δxl (i) on the left side in the x direction of the channel box 2b of the aggregate of interest and the water gap deformation amount δxr (i) on the right side in the x direction are equal (absolute values are equal). It is assumed that the sign is δx. Similarly, it is assumed that the water gap deformation amount δyt (i) on the upper side in the y direction of the target aggregate channel box 2b and the water gap deformation amount δyb on the lower side in the y direction are equal to δy. The changes in the aggregate core characteristics due to the deformation of the channel box are organized and referred to as a two-dimensional table or fitting formula in the form of v (δx, δy).

  Next, as a second step, correction is performed 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. In FIG. 11, the position of the channel box after 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 target aggregate channel box 2b is the deformation amount δxr (i−1) of the left side of the target aggregate channel box 2b and the right side of the left adjacent aggregate channel box 2a. The sum of The same applies to the water gap deformation amount δxr on the right side in the x direction.

  Since the deformation amounts of both adjacent aggregate channel boxes are generally different, if the deformation of both adjacent aggregate channel boxes is taken into consideration, δxl (i) ≠ δxr (i). Here, dx = δxl−δxr. Similarly, in the y direction, dy = δyt−δyb. Therefore, the table is first referred to as v (δxl, δyt), and the correction amount for this is arranged as w (dx, dy) and referred to. Finally, the nuclear property change is given by v + w.

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

(B) Single assembly neutron flux discontinuity factor diffusion model.
Next, method 2 will be described. In Method 2, the amount of change in the aggregate neutron flux discontinuity factor is calculated from the channel box deformation of the aggregate of interest and the adjacent aggregate, and the neutron flux discontinuity factor is calculated for the system in which each aggregate is homogenized. The neutron flux distribution in the assembly is obtained by the diffusion node method used, and the amount of change in the assembly nucleus properties is evaluated from this neutron flux distribution. 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 literature "Assembly Homogenization Techniques for Light Water Reactor Analysis," KSSmith, Progress in Nuclear Energy, vol. 17, p303, 1986. Has been. An example of a method for calculating the local power distribution in the assembly using the neutron flux distribution obtained by the node method calculation is "SIMULATE-3 Pin Power Reconstruction Methodology and Benchmarking," KRRempe et.al, Proceedings of International. Reactor Physics Conference, III-19, Jackson Hole, 1988.

  When the channel box is deformed, the aggregate neutron discontinuity factor changes because the neutron flux distribution in the aggregate changes due to the change in the water gap width. 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 the nodes where the aggregates are homogenized in the diffusion node method. That is, for example, the boundary condition of the equation (2) is given to the neutron flux at the boundary in the x direction between the target assembly n and the adjacent assembly m. Here, in Equation (2), f represents a neutron flux discontinuity factor at the boundary surface, and Ψ represents a neutron flux at the boundary surface.

  If the aggregate of interest and the adjacent aggregate are the same lattice type in an infinite lattice system, the neutron discontinuity factors are also equal, and the above boundary condition merely represents the continuity of the neutron flux. However, if the neutron discontinuity factor f on the adjacent surface between adjacent assemblies differs due to channel box deformation, a virtual source of neutrons is generated at the node boundary based on the above boundary conditions. Therefore, the change of neutron flux distribution in the aggregate due to channel box deformation can be calculated. In the multi-group diffusion node method, the nodal average thermal neutron flux and the nodal boundary thermal neutron flux are given by the whole core calculation for the thermal neutron flux using the neutron flux discontinuity factor.

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

  If the thermal neutron flux distribution in the homogenized assembly is obtained, the inhomogeneous thermal neutron flux distribution in the assembly can be calculated by synthesizing the non-homogeneous neutron flux distribution obtained by infinite lattice calculation. The assembly average nuclear constant, which is a function of the inhomogeneous neutron flux distribution, and the fuel rod local power distribution can be calculated immediately. The fuel rod R-factor distribution can be calculated from the fuel rod local power distribution.

  The change amount of the assembly neutron flux discontinuity factor with respect to the channel box deformation amount on each face of the assembly is prepared in advance as a table of channel box deformation amounts by single assembly detailed calculation (infinite lattice calculation). FIG. 14 shows an example of a functional form for the water gap width change δx of the assembly discontinuity factor with respect to the thermal neutron flux.

  As an example showing the calculation accuracy of this method, the channel box of the aggregate of interest is deformed by 2 mm in the x and y directions in the same way as in FIG. When there is no deformation, that is, when δx = δy = 2 mm, the local output distribution in the aggregate is calculated by this method, and the result compared with the reference solution by the multi-aggregate detailed calculation is as shown in FIG. It will be the same. The fuel rod local power calculation method was basically the same as that described in the above-mentioned K. Rempe document, but the correction of the inhomogeneity of the above-mentioned assembly with respect to changes in the thermal neutron flux distribution Was added. According to the reference calculation, the aggregate maximum local peaking coefficient changes by about 10% due to the channel box deformation, but this method reproduces this well.

  This method eliminates the need for multi-aggregate detailed calculations when preparing a table of changes in the aggregate neutron flux discontinuity factor from the channel box deformation, and the channel box deformation of the aggregate of interest and adjacent aggregates. When calculating the change of the assembly nucleus characteristics from the above, the neutron flux discontinuity factor of each surface of the fuel assembly may be calculated independently as a function of only the deformation amount of each surface of the channel box of the assembly of interest. As described above, the method 2 has an advantage that the preparation and reference of the table is simpler than the method 1 regardless of the combination of the displacement amounts of the respective surfaces.

  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 the infinite lattice system. When the spectral index is different between adjacent assemblies, thermal neutron spatial movement occurs, but the change from the infinite lattice of the thermal neutron flux is caused by a system consisting of nodes in which the target assembly and adjacent assemblies are homogenized. On the other hand, it is calculated by applying a diffusion model. In the modified first group node method, the 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 "Verification of LOGOS Nodal Method with Heterogeneous Burnup Calculations for a BWR core," T. lwamoto et al., Transaction of American Nuclear Society, vol. 71, p251, 1994.

  Even in the modified first group method, the thermal neutron flux change due to channel box deformation uses a neutron flux discontinuity factor as a boundary condition for the thermal neutron flux in a system consisting of nodes of homogenized aggregates of interest and adjacent aggregates. Thus, the calculation can be performed based on the same principle as in the above-described multigroup node method. Note that the thermal neutron flux and thermal neutron flow can also be calculated from the spectral indices of adjacent nodes using empirical weighting factors as is well known instead of solving the diffusion equation.

(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. In this method, using the difference in the discontinuity factor due to the difference in water gap width, the neutron flux and neutron flow at the boundary of the aggregate are obtained. Use the sensitivity factor to calculate the change in aggregate nuclear properties. As described above, this is because the diffusion node method using the neutron flux discontinuity factor pays attention to high accuracy for the neutron flux and neutron flow at the aggregate boundary.

  Based on the boundary perturbation method, an example of calculating local peaking in an aggregate using “neutron current / neutron flux” as the perturbation at the aggregate boundary is shown in the document “A Boundary Condition Ferturbation Method for Prediction of Pin Power Distribution in LightWater Reactors, "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) within the assembly for the fuel rod (x, y) is calculated from the above-described equation (8). The sensitivity coefficient F in the equation (8) is prepared in advance by single aggregate detailed calculation or multi-aggregate detailed calculation in which boundary conditions are changed.

  However, in the Rahnema literature, when the neutron current / neutron flux at the interface is calculated based on the diffusion theory, the aggregate neutron flux discontinuity factor is not used, The change in neutron flow / neutron flux due to local neutron flux distribution change near the water gap cannot be calculated.

In the present invention, the node boundary value r is a multi-group diffusion using a neutron flux discontinuity factor for a system consisting of nodes in which the aggregate of interest and the adjacent aggregate are homogenized, as in the diffusion node method described above. It can be obtained by applying the node method. Alternatively, in the modified first group difference method, the nodal boundary thermal neutron flux may be calculated from the spectral index of adjacent nodes 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 above-described equation (9).

  The above equation (8) shows the correction calculation method by the boundary perturbation method for the local output of the aggregate, but the correction calculation of other aggregate nuclear characteristic quantities is also performed for the neutron flow / neutron flux of each nuclear characteristic quantity. It can be calculated in a similar manner using the sensitivity coefficient.

(D) Analytical diffusion model.
Next, method 3 will be described. In Method 3, instead of using the diffusion node method in Method 2 to obtain the thermal neutron flux in the aggregate, the channel box deformations of the aggregate of interest and the adjacent aggregates are used to calculate the intra-aggregate based on the analytical model for the diffusion equation. The change in the thermal neutron flux distribution is obtained, and the amount of change in nuclear properties is evaluated from this thermal neutron flux distribution. In 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 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-described equation (1). In equation (1), x is the distance from the aggregate boundary, κ is the reciprocal of the thermal neutral pre-diffusion distance, and a is the water gap change width. The change in the thermal neutron flux distribution in the aggregate can be approximated by the product δΨ (x) δΨ (y) of the one-dimensional distribution in the x and y directions. Since Method 3 is based on the analysis model, there is an advantage that it is not necessary to prepare a table in advance.

  If combustion proceeds while the assembly is misaligned, the history effect of neutron spectrum change in the assembly due to the water gap width change and the history effect of the burnup distribution are accumulated, and the effect of instantaneous misalignment of the assembly The difference occurs. The phenomenon in which the neutron spectrum (ratio of thermal neutron flux to fast neutron flux) changes from the infinite system and causes an increase in uranium combustion delay and plutonium accumulation is called the spectral history effect. Further, the effect of causing the uranium combustion delay or the like due to the burnup distribution in the aggregate is called the single burn effect. Since these effects generally work in the direction of canceling out the instantaneous effects, it is necessary to appropriately consider these effects in order to avoid an excessive margin in the calculation of the thermal limit value.

  The amount of channel box deformation increases with combustion of the aggregate, and the spectral change due to channel deformation also changes with combustion. For this reason, when the combustion history effect of channel deformation is evaluated in the detailed combustion calculation by 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-aggregate calculation, and the spectral history effect and the single burn-up effect are corrected by the core performance calculation. Even in the diffusion node method using the discontinuous factor, the channel deformation effect can be calculated only as an instantaneous effect. Therefore, the spectral history effect and the single burn-up effect are corrected by the core performance calculation.

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

Therefore, in the present invention, this technique is applied to the spectrum history effect due to the channel box deformation, the burnup distribution and the spectrum history distribution in the fuel assembly corrected for the channel box deformation effect are calculated, and the channel history due to the combustion history effect due to the channel deformation. The correction amount such as the average nuclear constant of the assembly and the local distribution of fuel rods is calculated. For example, for the fuel rod local output, the spectrum history SH (x, y) at the fuel rod position (x, y) in the assembly is expressed by the above-described equation (10). In Eq. (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. The value in the lattice system is shown. The correction of the local output distribution LPF considering the spectrum history is given by the above-described equation (11) by using the spectrum history correction coefficient (∂Σ / ∂SH) for the thermal group fission cross section Σ.

  Similarly, the hysteresis effect caused by the burnup distribution occurring in the node due to the thermal neutron flux distribution change due to the channel deformation can be corrected using the sensitivity coefficient related to the burnup as shown in the above-mentioned document.

The block block diagram which shows the 1st Embodiment of this invention. Explanatory drawing which shows the combination pattern of D lattice bundle and an offset bundle. The characteristic view which shows an example of the burnup dependence of the local peaking coefficient in an offset bundle adjacency system. The characteristic view which shows the offset amount of a channel box, and the relationship of an assembly neutron flux discontinuity factor. Explanatory drawing which shows the effect of local output distribution calculation by the 1st Embodiment of this invention. The block block diagram which shows the 2nd Embodiment of this invention. Explanatory drawing which showed the evaluation model of the bending amount of the axial direction of the channel box by irradiation growth. The characteristic view which shows an example of the irradiation growth function by fast neutron irradiation. Explanatory drawing in the case of calculating | requiring the bending amount of the axial direction of the channel box by irradiation growth by the finite element method. The characteristic view which shows an example of the irradiation growth function by a burnup degree. Explanatory drawing which shows the relationship between the channel deformation amount and water gap deformation amount in the 2nd Embodiment of this invention. The characteristic view which shows the example of a change of the largest fuel rod local output with respect to the water gap width deformation | transformation amount (when (delta) x = deltay and (delta) y = 0) of an attention aggregate. The characteristic view which shows the example of a change of the maximum fuel rod local output with respect to the water gap width deformation | transformation amount (when dx = dy and dy = 0) of an adjacent assembly. The characteristic view which shows the example of a change of an assembly neutron flux discontinuity factor with respect to the water gap width deformation | transformation amount of an attention aggregate. FIG. 3 is a partially cutaway perspective view of a fuel assembly of a boiling water reactor. Explanatory drawing of the positional relationship of the water gap in the fuel assembly of a boiling water reactor. Explanatory drawing of D lattice bundle. Explanatory drawing which shows the state by which two bodies around the control rod were replaced by the offset bundle. Explanatory drawing which showed an example of the change of the local output distribution in the aggregate | assembly by offset bundle adjacency. Explanatory drawing which showed the bending of the axial direction of the channel box by fast neutron irradiation. Explanatory drawing which showed the positional relationship of the water gap by the deformation | transformation 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 Instrument tube 11 Assembly average nuclear constant calculation means 12 Reactor power distribution calculation means 13 Local power distribution calculation means 14 Line power calculation means 15 Limit power ratio Calculation means 16 In-furnace neutron flux measuring device count value calculation means 17 In-furnace 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 recalculating means 23 Line power density calculating means 24 Limit power ratio calculating means 25 In-core neutron flux measuring device count value calculating means 26 Control rod irradiation amount calculating means

Claims (8)

When calculating the core performance of a boiling water reactor with a D-lattice core loaded with a D-lattice bundle and an offset bundle, the critical eigenvalue, neutron flux distribution, and power distribution of the core are calculated based on the diffusion model using the nodal average nuclear constant. A means for calculating the power distribution in the reactor, a node boundary value calculating means for calculating the neutron flux and neutron flow at the node boundary surface homogenizing each aggregate based on the diffusion theory calculation using the neutron flux discontinuity factor, An infinite lattice corrected using the average power of the cross section of the assembly calculated by the power distribution calculating means in the reactor and the sensitivity coefficient of the local power of the fuel rod to the neutron flux and neutron flow at each surface of the aggregate boundary by the node boundary value calculating means A core performance calculation apparatus, comprising: a linear power density calculating means for calculating a linear power density of an assembly based on a local fuel rod power. When calculating the core performance of a boiling water reactor with a D-lattice core loaded with a D-lattice bundle and an offset bundle, the critical eigenvalue, neutron flux distribution, and power distribution of the core are calculated based on the diffusion model using the nodal average nuclear constant. A means for calculating the power distribution in the reactor, a node boundary value calculating means for calculating the neutron flux and neutron flow at the node boundary surface homogenizing each aggregate based on the diffusion theory calculation using the neutron flux discontinuity factor, An infinite lattice corrected using the average power of the cross section of the assembly calculated by the power distribution calculating means in the reactor and the sensitivity coefficient of the local power of the fuel rod to the neutron flux and neutron flow at each surface of the aggregate boundary by the node boundary value calculating means A linear power density calculating means for calculating the linear power density of the assembly based on the local output of the fuel rods, and an assembly cross-sectional plane calculated by the in-furnace power distribution calculating means. Core performance calculation, comprising: a limit power ratio calculation means for calculating a limit output ratio of the fuel assembly from the fuel rod R factor calculated using the power and the fuel rod local output by the linear power density calculation means apparatus. When calculating the core performance of a boiling water reactor with a D-lattice core loaded with a D-lattice bundle and an offset bundle, the critical eigenvalue, neutron flux distribution, and power distribution of the core are calculated based on the diffusion model using the nodal average nuclear constant. A means for calculating the power distribution in the reactor, a node boundary value calculating means for calculating the neutron flux and neutron flow at the node boundary surface homogenizing each aggregate based on the diffusion theory calculation using the neutron flux discontinuity factor, Infinite lattice corrected using the average power of the cross section of the assembly calculated by the power distribution calculating means in the reactor and the sensitivity coefficient of the fuel rod R factor with respect to the neutron flux and neutron flow at each face of the assembly boundary by the node boundary value calculating means A core performance calculation apparatus comprising: a limit power ratio calculation means for calculating a limit power ratio of the assembly by an R factor. When calculating the core performance of a boiling water reactor with a D-lattice core loaded with a D-lattice bundle and an offset bundle, the critical eigenvalue, neutron flux distribution, and power distribution of the core are calculated based on the diffusion model using the nodal average nuclear constant. A means for calculating the power distribution in the reactor, a node boundary value calculating means for calculating the neutron flux and neutron flow at the node boundary surface homogenizing each aggregate based on the diffusion theory calculation using the neutron flux discontinuity factor, Correction using the average cross-section power of the aggregate calculated by the power distribution calculation means in the reactor and the neutron flux and the neutron flux at each position of the aggregate boundary by the node boundary value calculation means and the sensitivity coefficient of the neutron flux at the position of the reactor in the reactor In-core neutron flux counter count value calculation means for calculating in-core neutron flux counter count and irradiation dose from the infinite lattice instrument position neutron flux Core performance computing device and butterflies. When calculating the core performance of a boiling water reactor with a D-lattice core loaded with a D-lattice bundle and an offset bundle, the critical eigenvalue, neutron flux distribution, and power distribution of the core are calculated based on the diffusion model using the nodal average nuclear constant. A means for calculating the power distribution in the reactor, a node boundary value calculating means for calculating the neutron flux and neutron flow at the node boundary surface homogenizing each aggregate based on the diffusion theory calculation using the neutron flux discontinuity factor, Infinite corrected by using the average power of the cross section of the aggregate calculated by the power distribution calculating means in the reactor and the sensitivity coefficient of the neutron flux at the control rod position with respect to the neutron flux and neutron flow at each face of the aggregate boundary by the node boundary value calculating means A core performance calculation apparatus comprising control rod irradiation amount calculation means for calculating a control rod irradiation amount from a lattice control rod position neutron flux. The in-core power distribution calculation means calculates the fast neutron flux distribution by the modified first group diffusion calculation, and uses the obtained fast neutron flux and the spectral index which is the ratio of the thermal neutron flux to the fast neutron flux in the infinite lattice calculation. 6. The core performance calculation apparatus according to claim 1, wherein a thermal neutron flux and a thermal neutron flow are calculated. The calculated value of the in-core power distribution is corrected by adapting the calculated value of the in-core neutron flux counter count value, which corrects the adjacent effect between the D lattice bundle and the offset bundle, to the actually measured value of the in-core neutron flux meter. 7. The core performance calculation apparatus according to claim 1, further comprising an in-core power distribution correction unit. For the fast neutron flux of the thermal neutron flux at the position of each fuel rod in the fuel assembly obtained based on the burnup distribution in the fuel assembly and the neutron flux and neutron flow on each face of the assembly boundary by the node boundary value calculation means Using the spectrum history distribution, which is the average value of the burnup ratio, calculate the correction amount for the node average nuclear constant, fuel rod local power distribution, and fuel rod R factor distribution due to the hysteresis effect associated with the adjacent D lattice bundle and offset bundle. The core performance calculation apparatus according to claim 1, wherein the core performance calculation apparatus according to claim 1 is used.