JP3441178B2 - Method and apparatus for calculating core performance of nuclear reactor - Google Patents

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a reactor core performance calculation method and a reactor core performance calculation apparatus.

[0002]

2. Description of the Related Art In the calculation of core performance in a nuclear reactor, a neutron diffusion equation or a transport equation based on a three-dimensional physical model of the core is solved by calculation to obtain an effective multiplication factor in the core,
Calculate neutron flux distribution and power distribution. The fuel rod power normalized by the cross section of the assembly is called the local power, and the thermal limit of the core is calculated by the average power of the cross section of the assembly and the local power of the fuel rod.

The calculation of the neutron flux distribution in the core is usually performed by dividing the core into a large number of small volumes (fuel nodes) each of which is a cube whose one side is the radial pitch of the fuel assembly. Solve the diffusion equation or the transport equation by approximating that the composition is homogeneous. In this case, the nuclear constant of the fuel node is generally given by a single-aggregate combustion calculation of a radial two-dimensional infinite system assuming that no neutron leaks into or leaks into the aggregate. The nuclear characteristics of the aggregate depend largely on the neutron spectrum distribution in the aggregate, that is, the distribution of the ratio of thermal neutron flux to fast neutron flux. Here, the neutron spectrum in the case of single-aggregate combustion calculation is called an asymptotic spectrum.

However, when the assembly adjoins another assembly having a different neutron spectrum in the core, neutrons move due to the spectrum interference effect with the adjacent assembly, and the neutron spectrum in the fuel node changes from the asymptotic spectrum. Change. When burned due to spectral interference, the nuclear constant of the fuel node also changes because the material composition in the fuel node changes from the case of single-assembly calculation. This is called the spectrum interference history effect.

Further, since the neutron flux distribution is distorted by the exchange of neutrons with the adjacent aggregates as compared with the single aggregate combustion calculation, the nuclear constant changes due to nonuniform combustion in the aggregates. The change in the nuclear constant due to non-uniform combustion is called the burnup distribution effect within the assembly. Further, the spectrum interference history effect and the burnup distribution effect are collectively referred to as the combustion history effect. Since the change in the axial fuel composition of the fuel assembly is gradual, the spectral interference effect between the axial nodes is smaller than the radial spectral interference effect. Therefore, the discussion of the spectrum interference effect below will be limited to the radial direction.

[0006]

One method of incorporating the combustion history effect is to calculate a nuclear constant by performing a combustion calculation in which a large number of aggregates are combined, and particularly in a boiling water reactor (BWR). In the case of, there is a problem that the number of combinations becomes enormous due to the void mismatch between the aggregates and the calculation cost is high. A method for assessing the history effect has been proposed by Lempe et al. (K.
Rempe, K.Smith and A.Henry, “SIMULATE-3 Pin Power R
econstruction Methodology and Benchmarking, ”Pro
ceedings of International Reactor Physics Conferen
ce, JacksonHole, USA, III-19, 1988).

In this evaluation method, in order to incorporate the spectrum interference history effect, the burnup average value (spectrum history) to the ratio of the asymptotic spectrum of the neutron spectrum of the fuel node is added as a parameter, and the deviation of the burnup average value is added. To change the homogenization nuclear constant. For example,
Change in thermal group fission cross section at fuel node position r δ
Σ _f2 (r) is given by the following equation.

[0008]

[Equation 9] The spectrum history SH (r) is an average value of the ratio of the neutron spectrum f at the position r in the fuel node to the asymptotic spectrum f∞ with respect to the burnup E, and is defined by the following equation.

[0009]

[Equation 10]

[0010]

[Outer 1]

However, in the evaluation method by Lempe et al., The neutron spectrum distribution in the fuel node is calculated by the ratio of the neutron flux of the heat group and the neutron flux of the high speed group obtained by the total core diffusion calculation of the neutron energy 2 group, so that calculation time is required. There is a problem.

Further, the spectrum history of the fuel node stores only the values for the four radial sides of the fuel node, and the distribution of the spectrum history in the fuel node is exp (-κx) times the value at the side of the fuel node. By asking.
Where κ is the reciprocal of the heat group diffusion distance of the fuel node,
x is the distance from one side of the fuel node to the point of interest in the fuel node. However, since this evaluation method is based on the one-dimensional model, there is a problem that the two-dimensional spectral distribution in the fuel node cannot be accurately considered.

As a method of incorporating the spectrum interference history effect, instead of adding the burnup average value of the ratio of the neutron spectrum of the fuel node to the asymptotic spectrum as a parameter, the burnup average value of the neutron spectrum of the fuel node itself is used. An operation planning method which is introduced as a parameter in place of the conventional hysteresis void fraction and arranges the nuclear constants by this parameter is disclosed in Japanese Patent Application Laid-Open No. 4-320996 (operation planning method for nuclear power plant and reactor and its planning device). ing.

However, this operation planning method is essentially the same as the evaluation method of Lempe et al., Only in the method of organizing the nuclear constants. In addition, JP-A-4-320996
As a method for evaluating the average neutron spectrum of a fuel node without calculating two groups (heat group and high-speed group), the publication discloses an asymptotic approach to the aggregate average of fuel nodes adjacent to a fuel node of interest. Although it is described that the weighted average value of the spectrum is taken, this weighting factor is empirically determined and there is a problem in accuracy. Further, this operation planning method has a problem that even if the neutron spectrum of the fuel node average can be evaluated, the neutron spectrum distribution in the fuel node cannot be evaluated.

Further, the neutron flux distribution is distorted by neutron exchange with adjacent aggregates (spectral interference effect) as compared with the single aggregate combustion calculation, and as a result, the nuclear constants are caused by nonuniform combustion in the aggregates. As for the change of, in the evaluation method of Lempe et al., The average value of the burnup on the sides of the fuel node and the volume average value of the fuel node are stored, and the burnup distribution in the fuel node is calculated based on these values. It is calculated by the following formula expansion.

However, as shown in the examples described later, since the spatial distributions of the burnup distribution due to the inclination of the high speed group and the burnup distribution due to the spectral interference are greatly different, these burnup distributions can be distinguished from each other. However, there is a problem in accuracy when the quadratic expansion is performed.

Further, the single-assembly combustion calculation for evaluating the nuclear constant of the fuel node is usually performed in a state where no control rod is inserted in the assembly. This is because it is difficult to previously specify the period in which the control rod is inserted for each fuel node. When control rods are inserted into the assembly, thermal neutrons are absorbed by the control rods, causing spectral hardening.

Therefore, a difference in nuclear constant occurs due to a difference in spectrum history between when the control rod is burned with the control rod inserted and when the control rod is instantly inserted by performing normal combustion. The difference in the nuclear constant due to the difference in the spectrum history is called the control rod history effect. The control rod history effect is a kind of spectrum history effect, but since the control rod is localized in the assembly, there is a large change in the local spectrum history within the assembly (single burn). . The effect of one-sided combustion becomes particularly remarkable with respect to the fuel rod local output, but the control rod history effect is not described in the Lempe evaluation method and the operation planning method described in Japanese Patent Laid-Open No. 4-320996.

As a method of correcting the control rod history effect for the local output of the BWR fuel rods, there is a method proposed in Japanese Patent Publication No. 2-11120 (reactor monitoring device). In this correction method, the normal local output peaking when there is no single burn is compared with the local output peaking of the control rod that is closest to the cross control rod considering the single burn effect, and the larger one is the maximum local output of the fuel node. It is the peaking. The one-sided combustion effect is given as a function of the period in which the control rod of the fuel node is inserted and burned, and the burning period after the control rod is pulled out.

However, the correction method described in Japanese Patent Publication No. 2-11120 does not take into consideration the effect on other nuclear constants in the fuel node, and the local output is the most centered on the cross control rod. There is a problem that the effect is considered only for the nearby fuel rods. Furthermore, there is a problem that the relationship with the spectrum interference history effect due to the exchange of neutrons with the adjacent aggregate is not taken into consideration. Incidentally, Sitterman et al. Have proposed a method of correcting the control rod hysteresis effect for infinite multiplication, which is one of the nuclear constants of fuel nodes (S. Sitarman and F. Rahnema, “Co.
ntrol BladeHistory Reactivity Model for Criticalit
y Calculations, ”Proceedings of Joint International
Conference on Mathematics and Supercomputing in Nu
clear Applications, p222, Karlsruhe, 1993).

In the method of correcting the control rod hysteresis effect by Sitterman et al., The infinite number of nodes is increased by the weighted average of the nuclear constants when the control rods are inserted and burned, and by the normal combustion calculation without the control rods. It is to obtain the magnification. However, in this method, there is a problem that the correction for the fuel rod output having a large one-sided combustion effect is not considered, and the spectral interference history effect due to the exchange of neutrons with the adjacent assembly is not considered. .

The present invention has been made in consideration of the above-mentioned circumstances, and corrects the combustion history effect by using the nuclear constant given by the single-assembly combustion calculation, and the power distribution in the core and the combustion in the nodes are corrected. An object of the present invention is to provide a core performance calculation method for a nuclear reactor and a core performance calculation device for the power distribution, which can accurately calculate the power distribution in a short time.

Another object of the present invention is to use a nuclear constant given by a single assembly combustion calculation, and output in the core by a spectrum interference history effect between adjacent assemblies, a burnup distribution effect, and a control rod history effect. The present invention provides a core performance calculation method for a nuclear reactor and a core performance calculation device capable of uniformly and accurately calculating distribution and correction of fuel rod power distribution in an assembly.

[0024]

In order to solve the above-mentioned problems, a nuclear reactor core performance calculation apparatus according to the present invention, as set forth in claim 1, incorporates a core state parameter from the core for fuel consumption. Core 3D neutron diffusion calculation device for calculating nuclear constants in nodes and calculation of fast neutron flux distribution in the core and asymptotic spectrum of fuel node, and fast neutron flux distribution and asymptotic spectrum from the core 3D neutron diffusion calculation device And a burnup E of the ratio of the spectral distribution from the spectral distribution calculation device to the asymptotic spectrum of the spectral distribution in the fuel node using the asymptotic spectrum. The mean value for, ie

[Equation 11] And an integrator that calculates the spectrum history SH (r) and the burnup distribution, and the spectrum history and burnup distribution from this integrator are used to determine the adjacency from the nuclear constant of the single-aggregate combustion calculation in the fuel node. Spectral history correction amount to correct the change of the nuclear constant due to neutron exchange with the fuel node and the change of the nuclear constant due to the non-uniform combustion from the nuclear constant of the single-aggregate combustion calculation in the fuel node A correction calculation device for calculating a burnup distribution correction amount, and a local output distribution calculation device for inputting the spectrum history correction amount and the burnup distribution correction amount to obtain a correction for a local output in the fuel node and calculating a local output distribution And the core three-dimensional neutron diffusion calculation device feeds back the spectrum history correction amount and the burnup distribution correction amount from the correction calculation device to perform a convergence calculation on the reactor. It is obtained to calculate the neutron flux distribution of the inner.

In order to solve the above-mentioned problems, the reactor core performance calculation device according to the present invention is characterized in that, as described in claim 2, the integration device calculates the spectral distribution in the fuel node and the asymptotic spectrum. The correction calculation device includes a spectrum history calculation device that inputs and calculates a spectrum history in the fuel node and a burnup distribution calculation device that calculates a burnup distribution. It is provided with a spectrum history calculation correction device for calculating a spectrum history correction amount for a number and a burnup distribution correction calculation device for calculating a burnup distribution correction amount.

In order to solve the above-mentioned problems, the reactor core performance calculation apparatus according to the present invention, as set forth in claim 3, incorporates the core state parameter from the core to determine the core in the fuel node. Reactor 3 which calculates the number and calculates fast neutron flux distribution in the core and asymptotic spectrum of fuel node 3
-Dimensional neutron diffusion calculation device, spectrum distribution calculation device for obtaining the spectrum distribution in the fuel node by inputting fast neutron flux distribution and asymptotic spectrum from this core three-dimensional neutron diffusion calculation device, and spectrum from this spectrum distribution calculation device Using the distribution, the average value for the burnup E of the ratio of the spectral distribution to the asymptotic spectrum in the fuel node, that is,

[Equation 12] And an integrator that calculates the spectrum history SH (r) and the burnup distribution, and the spectrum history and burnup distribution from this integrator are used to determine the adjacency from the nuclear constant of the single-aggregate combustion calculation in the fuel node. Spectral history correction amount to correct the change of the nuclear constant due to neutron exchange with the fuel node and the change of the nuclear constant due to the non-uniform combustion from the nuclear constant of the single-aggregate combustion calculation in the fuel node A correction calculation device that calculates the burnup distribution correction amount, and a control within the fuel node that is the spectrum history within the fuel node obtained from the spectral distribution within the fuel node during the control rod insertion period and its withdrawal period A control rod history calculator that calculates rod history and a control rod history from this control rod history calculator are input to calculate the correction amount for the nuclear constant of the fuel node. And a control rod history correction amount from the correction calculation device and a control rod history correction amount from the control rod history correction calculation device are input to the local output in the fuel node. And a local power distribution calculation device for calculating a local power distribution by obtaining a correction, wherein the core three-dimensional neutron diffusion calculation device is configured to calculate the spectrum history correction amount and burnup distribution correction amount and control rod history correction calculation from the correction calculation device. The control rod history correction amount from the device is fed back to calculate the neutron flux distribution in the core by convergence calculation.

On the other hand, in order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention divides the core of the nuclear reactor into small volumes and sets it into a single set, as described in claim 4. Reactor core performance calculation method for calculating the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by body burning calculation In, the spectral distribution, which is the distribution of the ratio of thermal neutron flux to fast neutron flux in the small volume, is obtained by solving the diffusion equation for the small area in the core consisting of the small volume and the adjacent area surrounding it, The average value for the burnup E of the ratio of the small volume spectral distribution to the asymptotic spectrum using the spectral distribution, that is,

[Equation 13] From the single-unit combustion calculation of the nuclear constants due to the fact that the small volume burns in response to neutron exchange with the adjacent small volume using this spectrum history Is a method of obtaining the change of.

In order to solve the above-mentioned problems, the method for calculating the core performance of a nuclear reactor according to the present invention is, as described in claim 5, the distribution of the ratio of thermal neutrons to fast neutrons in a small volume, It is a method to obtain non-separable type expansion by an analytical solution obtained by analytically solving the neutron diffusion equation with the boundary value being the ratio value on the side of a small volume.

Furthermore, in order to solve the above-mentioned problems, the method for calculating the core performance of a nuclear reactor according to the present invention divides the core of the nuclear reactor into small volumes to form a single set as described in claim 6. Reactor core performance calculation method for calculating the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by body burning calculation In, the spectral distribution, which is the distribution of the ratio of thermal neutron flux to fast neutron flux in the small volume, is obtained by solving the diffusion equation for the small area in the core consisting of the small volume and the adjacent area surrounding it, The average value for the burnup E of the ratio of the small volume spectral distribution to the asymptotic spectrum using the spectral distribution, that is,

[Equation 14] The spectrum history SH (r) is obtained by non-separable expansion by an analytical solution of the neutron diffusion equation whose boundary value is a value on the side of the small volume, and the small volume is adjacent to the small volume by using this spectrum history. It is a method of obtaining the change from the single-aggregate combustion calculation of the nuclear constant due to the combustion due to the interaction of neutrons with the small volume.

In order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention divides the reactor into small volumes of the core of the nuclear reactor to form a single set. Reactor core performance calculation method for calculating the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by body burning calculation In, in the interaction of neutrons with the adjacent small volume, the change from the single-aggregate combustion calculation of the nuclear constant due to the fact that the small volume burned by receiving the interaction of neutrons with the adjacent small volume, Based on the distribution of burnup within the small volume, it is divided into a component based on the gradient of the fast neutron flux and a component based on the distortion of the spectrum, and the component based on the gradient of the fast neutron flux is the side of the small volume. Is obtained by the bi-multistage polynomial expansion whose boundary value is the value at, and the component based on the distortion of the spectrum is obtained by the non-separable expansion of the neutron diffusion equation whose boundary value is the value on the side of the small volume. Is the way.

Furthermore, in order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention divides the core of the nuclear reactor into small volumes as described in claim 8,
Reactor core that calculates the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by the single-assembly combustion calculation In the performance calculation method, the spectrum which is the ratio of thermal neutron flux to fast neutron flux in the small volume is obtained by solving the diffusion equation for the small area in the core consisting of the small volume and the adjacent area surrounding it, Using the spectrum, the average value of the ratio of the small volume spectrum to the asymptotic spectrum to the burnup E, that is,

[Equation 15] Is obtained, the change from the single-aggregate combustion calculation of the nuclear constant due to the small volume burned by receiving neutron interaction with the adjacent small volume, the spectrum within the small volume The method is characterized in that the sensitivity coefficient for the spectral history is obtained by using a history and is a function of an average value for the burnup of the small volume and the burnup of the water density.

On the other hand, in order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention divides the core of the nuclear reactor into small volumes to form a single set as described in claim 9. Reactor core performance calculation method for calculating the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by body burning calculation In, a spectrum that is the ratio of thermal neutron flux to fast neutron flux in the small volume, obtained by solving the diffusion equation for the small region in the core consisting of the small volume and adjacent regions surrounding it, using this spectrum The average value of the ratio of the small volume spectrum to the asymptotic spectrum to the burnup E, that is,

[Equation 16] Is obtained, the change from the single-aggregate combustion calculation of the nuclear constant due to the small volume burned by receiving neutron interaction with the adjacent small volume, the spectrum within the small volume The sensitivity coefficient for the spectrum history is obtained by using a history, and is a method for obtaining the sensitivity coefficient for the small volume by an aggregate calculation in consideration of leakage or leakage of neutrons.

In order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention divides the core of the nuclear reactor into small volumes to form a single assembly. Reactor core performance calculation method for calculating the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by body burning calculation In the above, the small volume is obtained by calculating the change from the single-aggregate combustion calculation of the nuclear constant due to the burning due to the interaction of neutrons with the adjacent small volume, and the nuclear constant used when obtaining this change. The sensitivity coefficient for burnup is obtained by unit cell burnup calculation consisting of only one fuel rod, or from the burnup dependence of the advanced burnup point with respect to the aggregate mean nuclear constant. A method of obtaining the polynomial approximation for burnup.

Further, a method for calculating core performance of a nuclear reactor according to the present invention is intended to solve the above-mentioned problems.
, The nuclear core of the reactor is divided into small volumes, and the neutron diffusion equation or transport equation is solved by using the nuclear constant for the small volume given by the single assembly combustion calculation to solve the neutrons in the core. In a reactor core performance calculation method for calculating a bundle distribution and a fuel rod power distribution in an assembly, a single nuclear constant resulting from the fact that the small volume burns in response to neutron exchange with an adjacent small volume. Obtaining the change from the aggregate burnup calculation, the burnup coefficient of the nuclear constant and the sensitivity coefficient for the burnup average value of the ratio of the thermal neutron flux to the fast neutron flux used when determining this change are:
This is a method of separately providing a fuel rod containing no burnable poison and a fuel rod containing burnable poison.

Furthermore, in order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention divides the core of the nuclear reactor into small volumes to achieve Calculation of the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or transport equation using the nuclear constant for the small volume given by the assembly combustion calculation. In the method, a change in nuclear constant due to burning of a control rod inserted in the small volume from a single-aggregate combustion calculation is calculated as a non-homogeneous calculation of the aggregate when the control rod is inserted in the small volume. The spectral distribution, which is the distribution of the ratio of thermal neutron flux to fast neutron flux in the small volume given by, is obtained, and using this spectral distribution, the asymptotic spectrum of the spectrum of the small volume is calculated. Mean values for burnup E a ratio le, i.e.,

[Equation 17] This is a method of obtaining a spectrum history SH (r) that is and correcting using this spectrum history.

[0036]

In order to solve the above-mentioned problems, the core performance calculation method for a nuclear reactor according to the present invention is, as described in claim 13, characterized in that the small volume resulting from the insertion of the control rod into the small volume is reduced. This is a method of obtaining the amount of change in the nuclear constant within the volume from the single aggregate combustion calculation using a correction coefficient according to the distance from the control rod.

[0038]

In order to solve the above-mentioned problems, the method for calculating the core performance of a nuclear reactor according to the present invention, as set forth in claim 14, has several fuels which are thermally severe in a small region of the core. Select the bar as a candidate and for this candidate,
This is a method of calculating the fuel rod power in consideration of the change in the nuclear constant from the single-assembly burnup calculation.

Further, in order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention divides the core of the nuclear reactor into small volumes to form a single set, as described in claim 15. Reactor core performance calculation method for calculating the neutron flux distribution in the core and the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by body burning calculation In, the spectral distribution, which is the distribution of the ratio of thermal neutron flux to fast neutron flux in the small volume, is obtained by solving the diffusion equation for the small area in the core consisting of the small volume and the adjacent area surrounding it, The average value for the burnup E of the ratio of the small volume spectral distribution to the asymptotic spectrum using the spectral distribution, that is,

[Equation 18] From the single-aggregate combustion calculation of the nuclear constant due to the fact that the small volume burns in response to the exchange of neutrons with the adjacent small volume using this spectrum history SH (r) Is calculated, and the limit output ratio, which is a monitoring index for burnout of the fuel rod, is calculated by using the fuel rod output in consideration of this change.

Further, in order to solve the above-mentioned problems, the method of calculating the core performance of a nuclear reactor according to the present invention sets forth a claim 16.
In addition to the change from the single-aggregate combustion calculation of the nuclear constant due to the small volume burned in response to the interaction of neutrons with the adjacent small volume, By using the fuel rod output that takes into account the change from the single-assembly combustion calculation of the nuclear constant due to the burning and insertion of is there.

[0042]

According to the reactor core performance calculation method and apparatus according to the present invention, with the above-described configuration, (1) the spatial distribution of the spectrum in the fuel node is calculated by using a known modified first group model. The calculation time can be shortened by deciding only in the neighborhood. (2) Spectral history can be obtained by performing non-separable type expansion by the analytical solution of the diffusion equation for the spectral distribution and spectral history distribution in the node of interest. The accuracy of the burnup distribution effect can be improved. (3) The burnup distribution in the fuel node can be separated into the contribution of the inclination of the high speed group and the contribution of the distortion of the spectrum to expand the accuracy of the burnup distribution effect. (4) It is possible to improve the calculation accuracy of the sensitivity coefficient for the burnup distribution of the nuclear constant and the spectrum history distribution, and (5) the effect of the spectrum interference history. It is possible to combine the effect of the control rod history.

As a result, it is possible to accurately calculate the power distribution (neutron flux distribution) in the core and the fuel rod power distribution in the fuel node.

In the reactor core performance calculation device according to the first aspect of the present invention, even if the assembly burns due to neutron exchange with another assembly having a different spectrum in the core, the single assembly is burned. Correct the combustion history effect using only the nuclear constant given by the combustion calculation, and in a sufficiently short calculation time,
It is possible to accurately calculate the power distribution in the core and the fuel rod power distribution in the assembly.

In the reactor core performance calculation device according to the second aspect, the spectral distribution and burnup distribution in the fuel node are obtained using an integrating device, and the spectral distribution and burnup distribution are corrected by the correction calculation device. Therefore, the spectral distribution and burnup distribution in the fuel node can be accurately obtained, and the power distribution in the core and the fuel rod power distribution in the fuel node can be quickly and accurately obtained.

According to the reactor core performance calculation device of the third aspect, the assembly burns by receiving neutron exchange with another assembly having a different spectrum in the core,
Even if a control rod is inserted into the assembly and burns, the combustion history effect is corrected using only the nuclear constants given by the single assembly combustion calculation, and the power distribution in the core is accurately calculated in a sufficiently short calculation time. And the fuel rod power distribution in the assembly can be calculated.

In the method for calculating core performance of a nuclear reactor according to claim 4, the neutron spectrum is close to the asymptotic value in the single-assembly combustion calculation at the center of the fuel node (small volume) of the assembly size. Then, the neutron spectrum in the node of interest can be calculated by solving the diffusion equation in a small region in the core by the spectrum distribution calculation device, which enables the fuel node to be calculated without performing the total core calculation for thermal neutrons. The spectrum history inside can be obtained accurately, and the calculation time can be shortened significantly.

In the method of calculating core performance of a nuclear reactor according to claim 5, the spectral distribution calculation device is used to calculate the spectral distribution in the fuel node (small volume) by the neutron spectrum at a point on the side of the fuel node. By analytically solving the diffusion equation with the value of as the boundary value and obtaining it by the non-decomposition type expansion, it is possible to further reduce the time for solving the diffusion equation and calculate the spectrum history.

In the reactor core performance calculation method according to the sixth aspect, the expansion formula for the spectrum distribution in the fuel node (small volume) is applied to the spectrum history distribution, and the spectrum history calculation device is used to The spectrum history is obtained by the non-separable expansion based on the values on the sides, and the accuracy of the spectrum history distribution is improved.

In the reactor core performance calculation method according to the seventh aspect, the contribution of the inclination of the high speed group and the contribution of the spectrum interference to the burnup distribution in the fuel node due to the exchange of neutrons with the adjacent node are separated. By doing so, the burnup distribution in the fuel node can be accurately calculated by the burnup distribution calculation device, and the correction accuracy of the nuclear constant due to the burnup distribution in the fuel node can be improved. is there.

In the reactor core performance calculation method according to claim 8, the sensitivity coefficient for the spectral history of the nuclear constant used to correct the spectral interference history due to the exchange of neutrons with the adjacent node, and the historical water density The sensitivity coefficient is improved by expressing the sensitivity coefficient as a function of the burnup and the history water density or the history void ratio by the spectral history correction calculation device in consideration of the dependency of the above.

In the core performance calculation method for a nuclear reactor according to claim 9, the sensitivity coefficient is calculated by a spectrum history correction calculation device by an aggregate combustion calculation considering leakage or leakage of neutrons (neutrons with adjacent nodes). It is used to correct the spectrum interference history due to the exchange of)) and the accuracy of the sensitivity coefficient for the spectrum history of the nuclear constant is improved.

In the core performance calculation method for a nuclear reactor according to claim 10, the burnup distribution correction of the burnup dependence of the nuclear constant of the burnup distribution in the fuel node due to the exchange of neutrons with the adjacent leads is performed. Calculated by unit cell combustion calculation using a calculation device, or by polynomial approximation (quadratic fit) regarding burnup from burnup dependency at the point where burnup progressed with respect to the aggregate constant The accuracy of the burnup dependency of the nuclear constant used to correct the change of is improved.

In the core performance calculation method for a nuclear reactor according to claim 11, the sensitivity coefficient for the burnup of the nuclear constant used for correction and the average value of the burnup is calculated by using a burnup distribution correction calculation device as an ordinary fuel rod. It is given separately with fuel rods containing burnable poisons to improve the accuracy of the nuclear constant used for correction by the spectrum history and burnup distribution in the fuel node due to the exchange of neutrons with adjacent nodes. .

In the core performance calculation method for a nuclear reactor according to claim 12, when the control rod is inserted, which is given by non-homogeneous detailed calculation, in order to correctly evaluate the spatial distribution of the spectrum history due to the control rod insertion. By using the distribution of the ratio of the thermal neutron flux to the fast neutron flux in the assembly of, the control rod history calculation device calculates the spectrum history in the control rod insertion node to determine the control rod history effect as the spectrum history sensitivity coefficient. The control rod hysteresis effect can be treated as a spectrum hysteresis effect in a unified manner.

[0056]

In the reactor core performance calculation method according to the thirteenth aspect, the spatial distribution of the control rod hysteresis effect is taken into consideration, and the control rod hysteresis correction calculation device is used to calculate the nuclear constant based on the control rod hysteresis effect. The change from the one-bundle combustion calculation is calculated in advance and corrected as a function of the distance from the control rod,
The control rod history effect is directly corrected.

[0058]

In the method of calculating core performance of a nuclear reactor according to claim 14, when correcting a change in fuel rod output in a node caused by burning due to exchange of neutrons with an adjacent node, local correction is performed. Using the power distribution calculator, select a few fuel rods that are predicted to be the most thermally demanding, and consider the change from the single-assembly calculation of nuclear constants for this candidate. By calculating the local power peaking described above, the time required to calculate the fuel rod power can be significantly shortened. This can be suitably applied to online core performance monitoring.

In the reactor core performance calculation method according to the fifteenth aspect, the fuel rod output in consideration of the change in the nuclear constant in the fuel node due to the combustion due to the exchange of neutrons with the adjacent node. The calculation accuracy of the limit output ratio can be improved by calculating the limit output ratio, which is a monitoring index of burnout of the fuel rods, by using the local output distribution calculation device.

In the reactor core performance calculation method according to the sixteenth aspect, the fuel rod output is used in consideration of the change in the nuclear constant in the fuel node due to the combustion of the control rod inserted in the assembly. Then, by calculating the R factor with the local power distribution calculation device, it is possible to accurately calculate the limit power ratio, which is a fuel rod burnout monitoring index.

As described above, according to the core performance calculation method and apparatus of the present invention, (1) the spectral history distribution in the fuel node is used only in the vicinity of the node of interest, and the non-separable expansion by the analytical solution of the diffusion equation is used. Therefore, the accuracy of the spectrum history effect is improved. (2) Regarding the burnup distribution effect, the burnup distribution in the fuel node can be calculated separately for the contribution of the inclination of the high speed group and the contribution of the spectrum interference, and the calculation accuracy of the burnup distribution in the fuel node is improved. (3) Regarding the control rod history effect, the spectrum history effect is calculated in consideration of the locality of the effect.

Therefore, it is possible to correct the spectrum interference history effect, burnup distribution effect, and control rod history effect with respect to the power distribution in the core and the fuel rod power distribution in the fuel node accurately in a short time.

[0064]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described below with reference to the accompanying drawings.

In explaining the core performance calculation of the reactor according to the present invention, the basic concept of the core performance calculation of this reactor will be described.

In the core performance calculation of this reactor, the core of the reactor is divided into small volumes (fuel nodes), and a known modified one group model is used to determine the spatial distribution of the neutron spectrum in this fuel node. To do. Then, based on this modified 1-group model, the diffusion equation is solved only in the vicinity of the node of interest to obtain the distribution of the neutron spectrum (ratio of the thermal group neutron flux to the fast group neutron flux) in the fuel node. Compute the spectral history distribution and burnup distribution of. The spectral history distribution and burnup distribution may be calculated numerically, but in order to further shorten the calculation time, the spectral distribution is represented by a non-separable expansion using an analytical solution of the diffusion equation.

For details of the method for determining the spatial distribution of the neutron spectrum in the fuel node, see Iwamoto et al. (T. Iwam
oto and M. Tsuki, “New Nodal Diffusion and Pin Po
werCalculationd Method Based on Modified One Group
Scheme, ”Proceedings of the Topical Meeting on Ad
vance in Reactor Physics, Charleston, USA, pl-476,1
992). The neutron spectrum f (r) at the position r in the fuel node is expressed by the following equation.

[0068]

[Formula 19] The subscript j represents the four sides of the fuel node (see FIG. 2), k represents the four vertices of the fuel node, and j1 and j2 represent the two sides sandwiching the vertex k. Expansion coefficient C
j (C1 to 4) and Ck (C5 to 8) are determined by the neutron spectrum at points (indicated by white circles in FIG. 2) on the sides of the fuel node.

The distribution of the spectrum history is obtained by averaging the burnup ratio of the neutron spectrum distribution in the fuel node and the asymptotic spectrum. In this case as well, only the spectrum history on the sides of the fuel node is stored and the fuel node is stored. The neutron spectrum distribution in can be expressed by expansion with the non-separable type by the analytical solution similar to equation (3). Furthermore, based on the spectrum history distribution in the fuel node, the component based on the distortion of the neutron spectrum due to the exchange of neutrons (spectral interference effect) with the adjacent node in the burnup distribution in the fuel node can be calculated.

The control rod history effect is a kind of spectrum history effect, but since the spectrum history effect has a large locality in the fuel node, the spectrum history effect is taken into consideration and combined with the spectrum interference history effect. .

FIG. 1 is a block diagram showing a first embodiment of core performance calculation of a nuclear reactor according to the present invention.

The core performance calculation device 10 is operated by the core three-dimensional neutron diffusion calculation device 12 for inputting core state parameters which are output data from the core 11 of the nuclear reactor, and the core three-dimensional neutron diffusion calculation device 12. The fuel node spectral distribution calculation device 13 for inputting the fast neutron flux distribution and the asymptotic spectrum, the integrating device 14 for inputting the fuel node spectral distribution calculated by the spectral distribution calculating device 13, and the integrating device 14 for calculation Correction calculator 1 for inputting the recorded spectrum history and burnup distribution
5 and a fuel node local power distribution calculation device 16.

Of these, the core three-dimensional neutron diffusion calculation device 12 takes in core state parameters from the core 11,
The nuclear constant in the fuel node is calculated, and the fast neutron flux distribution in the core and the asymptotic spectrum of the fuel node are calculated based on the known modified one-group diffusion calculation.

In addition, the fuel node spectrum calculation apparatus 1
3 receives the fast neutron flux distribution and asymptotic spectrum from the core three-dimensional neutron diffusion calculation device 12, and analytically obtains the spectral distribution in the fuel node.

The obtained fuel node spectral distribution is input to the integrating device 14. The integrating device 14 includes a fuel node spectrum history calculation device 18 and a fuel node burnup distribution calculation device 19, which inputs and calculates the spectrum distribution in the fuel node to calculate the spectrum history and combustion in the fuel node. It is designed to calculate the degree distribution.

On the other hand, the correction calculation device 16 includes a nuclear constant spectrum history correction calculation device 20 in the fuel node and a nuclear constant burnup distribution correction calculation device 21 in the fuel node. The burnup distribution is input and the spectrum history correction amount and the burnup distribution correction amount for the fuel node core constant are output.

The calculation results of the spectrum history correction amount and the burnup distribution correction amount are fed back to the three-dimensional neutron diffusion calculation device 11 and the neutron flux distribution in the core 10 is calculated by the convergence calculation.

Further, the fuel node local output distribution calculation device 16 inputs the spectrum history correction amount and burnup distribution correction amount with respect to the fuel node core constant by the spectrum history correction device 15 to determine the local output distribution in the fuel node. It is supposed to calculate.

By the way, the reactor core performance calculation device 10
In-fuel-node spectral distribution calculation device 13
Is the fuel node of interest (hereinafter referred to as the node of interest)
The neutron spectrum within 23 is calculated by solving the diffusion equation in a small region within the core 11 consisting of eight fuel nodes 24 surrounding the node of interest 23 as shown in FIG. This is because the diffusion distance of thermal neutrons is smaller than the fuel node width, and at the center of the fuel node of the assembly size,
The neutron spectrum takes advantage of the property of approaching the asymptotic value in the single-assembly combustion calculation, and can be solved by numerical calculation by giving the asymptotic value of the neutron spectrum as a boundary condition at the center of each fuel node. As a result, the spectrum history in the fuel node can be accurately obtained without performing the whole core calculation for thermal neutrons, and the calculation time can be greatly shortened.

Next, a modification of the intra-fuel node spectral distribution calculation device 13 will be described.

The spectral distribution calculation device 13 in the fuel node shown in this modification further shortens the time for solving the diffusion equation, and calculates the neutron spectral distribution in the fuel node as
The neutron spectrum value at a point on the side of the fuel node 24 indicated by the white circle in FIG. 2 is used as a boundary value, and is expressed by a non-separable expansion based on the analytical solution of the diffusion equation expressed by the equation (3). Compute the spectral history in.
The neutron spectrum at the white circles is analytically given by the known modified one-group model as shown in the above-mentioned reference by Iwamoto et al.

The fuel-node spectrum history calculation device 18 for inputting and calculating the fuel-node spectrum distribution
Improves the accuracy of the spectral history within the fuel node. In the conventional evaluation method of Lempe et al., The accuracy is insufficient because the spectrum history inside the fuel node is developed one-dimensionally from the value of the spectrum history around the fuel node. However, this spectrum history calculation device 18
Then, the expansion formula (3) for the spectral distribution in the fuel node is also applied to the spectrum history, and the value is obtained by the non-separable expansion from the values on the sides of the fuel node.

The spectral history SH (r) at position r in the fuel node is

[Equation 20] d _{1 to 8} are on the sides of the fuel node 24 shown by the white circles in FIG.
Can be determined by the spectrum history at
It is a history coefficient. With this history coefficient d1 to 8,
Calculation accuracy for the two-dimensional distribution of the spectrum history in
It is possible to improve. This spectral distribution calculator
As an example showing the effect of No. 13, high concentration surrounded by low enriched fuel
About the deviation from 1 of the spectrum history in condensed fuel,
Example of calculation by the spectrum history calculation device 18 (black circle)
Fig. 3 shows a comparison between and the detailed calculation (white circles).

Next, the in-fuel node burnup distribution calculation device 19 for receiving and calculating the spectral distribution in the fuel node will be described.

The burnup distribution calculation device 19 provides a method for improving the correction accuracy of the nuclear constant due to the burnup distribution in the node due to the exchange of neutrons with the adjacent node.

In the evaluation method of Lempe et al., Only the average value of the burnup on the sides of the fuel node and the volume average value of the fuel nodes are stored. It is calculated by biquadratic expansion based on the value. Because of this, the component due to the spatial distortion of the spectrum cannot be represented correctly.

As an example, the change in the fuel rod local power distribution due to the instantaneous spectrum interference at a certain fuel node from the single-assembly combustion calculation, and the contribution of the spectrum interference and the slope of the fast neutron flux in the change. Is shown in FIG. As can be seen, the contribution of the spectral interference changes exponentially near the surface of the fuel node, while the contribution of the fast group changes gently throughout the fuel node.

In this embodiment, the burnup distribution in the fuel node can be accurately calculated by separating the contribution of the inclination of the high speed group and the contribution of the spectrum interference to the burnup distribution. That is, the component based on the gradient of the fast neutron flux is obtained from the burnup on the side of the fuel node by bi-polynomial expansion, and the component based on the distortion of the spectrum is calculated from the value on the side of the fuel node by the analysis function similar to the spectrum history. It is represented by the separation type expansion.

The deviation 8 (r) of the burnup distribution in the fuel node from the node average burnup E is the component δE _SH (r) based on the spectral interference and the component δE based on the slope of the fast neutron flux.
It is given by the sum of _FG (r). Node average burnup E of burnup distribution in fuel node due to components based on spectral interference
The deviation δE _SH (r) from is expressed by the following equation.

[0090]

[Equation 21] The deviation δE _SH (r) of the spectrum interference component from the node average burnup E of the burnup distribution in the fuel node can be obtained from the equation (2), and the integral of the second term on the right side of the equation (5) is the fuel. The integration in the in-channel region of the assembly is shown, and thus the second term on the right side represents the deviation from 1 of the average spectral history of the fuel rod portion of the assembly.

[0091]

[Equation 22] The expansion coefficient g is calculated from the value at the four midpoints of the fuel node shown in FIG.

The fuel node nuclear constant spectrum history correction calculation device 20 as the correction calculation device 15 provided in the reactor core performance calculation device 10 will be described. The spectrum history correction calculation device 20 relates to improvement of the accuracy of the sensitivity coefficient for the spectrum history of the nuclear constant used for correcting the spectrum interference history due to the exchange of neutrons (spectral interference effect) with the adjacent node.

This spectrum history correction calculation device 20 is
The sensitivity coefficient is a function of burnup and hysteresis water density or hysteresis void ratio. FIG. 5 and FIG. 6 show burning changes of the sensitivity coefficient with respect to the spectrum history of the infinite multiplication factor and the thermal group fission cross section, respectively. Than this,
It can be seen that the sensitivity coefficient has a dependency on the hysteresis void ratio, that is, the hysteresis water density.

[0094]

[Outside 2]

[0095]

[Equation 23]

Further, the following modification is also conceivable as the nuclear constant spectrum history correction device 20 in the fuel node. The spectrum history correction apparatus 20 of this modified example relates to improvement of accuracy of sensitivity coefficient with respect to spectrum history of nuclear constant used for correction of spectrum interference history due to interaction of neutrons with adjacent node (spectral interference effect), The sensitivity coefficient is obtained by an aggregate combustion calculation that considers the leakage or leakage of neutrons.

In the evaluation method of Lempe et al., The sensitivity coefficient is extracted from the single-assembly combustion calculation in which the neutron spectrum is changed by changing the hysteresis void ratio. However,
The spectral change of neutrons due to the change of void fraction is almost uniform in the aggregate, while the spectral change due to the exchange of neutrons with adjacent aggregates (spectral interference effect) is as shown in FIG. There is a big difference that it is large on the surface.

Therefore, in a normal aggregate having an enrichment distribution or a concentration distribution of burnable poisons in the aggregate, a difference in sensitivity coefficient occurs depending on the method of changing the spectrum. As an example, a comparison between the sensitivity coefficient obtained by the single-assembly combustion calculation in which the neutron leakage, which is considered to be closer to reality, is given by the Albet boundary condition, and the sensitivity coefficient extracted by the combustion calculation in which the void history is changed is shown in FIG. Shown in. Here, the triangle mark is the sensitivity coefficient by the albeto calculation. From this, it can be seen that there is a considerable error in the sensitivity coefficient extracted by the whid history.

Further, the correction calculation device 15 provided in the reactor core performance calculation device 10 has a nuclear constant burnup distribution correction calculation device 21 in the fuel node. This burnup distribution correction calculation device 21 relates to improvement in accuracy of burnup dependency of a nuclear constant used for correction of change in nuclear constant due to burnup distribution in a fuel node due to exchange of neutrons with an adjacent node. is there. In this example, the burnup dependence of the nuclear constant is
Whether to obtain by unit cell combustion calculation consisting of one fuel rod,
Alternatively, it is obtained by a quadratic fit regarding the burnup from the burnup dependency at the point where the burnup progresses with respect to the aggregate average constant.

The white circles in FIG. 8 indicate a fuel assembly.
The burn-up dependency of the cross-sectional area of the heat group nucleus separation of the aggregate average is shown for the hysteresis void fraction (VH) of 0%, 40%, and 70%. On the other hand, the heat group nuclear fission cross section fitted with a quadratic equation at a burnup of 10 GWd / st or more is also shown.

Although the heat group fission cross section for a fuel rod is originally a monotonically decreasing function of burnup,
The reason why the aggregated average value of white circles is not so is due to the distribution of concentration and burnable poison in the aggregate. When the average nuclear constant is used as described above, the burnup distribution effect on the fuel rod output cannot be corrected correctly, but can be correctly evaluated by using the method of this embodiment.

Next, a modification of the burnup distribution correction calculation device 21 in the combustion node provided in the reactor core performance calculation device 10 will be described.

The burnup distribution correction calculation device 21 relates to the improvement of the accuracy of the nuclear constant used for the correction by the spectrum history in the node and the burnup distribution due to the exchange of neutrons with the adjacent node. In this embodiment, the nuclear constants used for correction are separately provided for a normal fuel rod having no burnable poison and a fuel rod containing burnable poison. The reason why the nuclear constants used for correction are given separately is that the burning characteristics of the nuclear constants of fuel rods containing burnable poisons are significantly different from those of ordinary fuel rods. This improves the accuracy of the fuel rod power distribution in particular.

Further, the core performance calculation device 10 incorporates a fuel node local power distribution calculation device 16. This local power distribution calculation device 16 is a method for correcting a change in fuel rod power in a fuel node due to combustion caused by neutron exchange with an adjacent fuel node,
The correction for the local output of the fuel rod is performed by using the sensitivity coefficient for the burnup and spectrum history of the local output of each fuel rod. In this embodiment, the local output distribution can be directly corrected easily without using the sensitivity coefficient for the nuclear constant such as the thermal group fission cross section.

The fuel node local distribution calculation device 16 may be configured as follows.

This local power distribution calculation device 16 is the most thermally strict in correcting the change in the fuel rod output in the fuel node due to the burning due to the exchange of neutrons with the adjacent fuel node. Several fuel rods that are expected to be selected are selected as candidates, and local power peaking is calculated for these candidates in consideration of the change from the single-assembly combustion calculation of the nuclear constant.

The local output peaking coefficient p (r) when the nuclear fission cross section Σ _f2 of the heat group of the fuel rod r of _interest changes due to the effect of the combustion history is:

[Equation 24] The integral of the denominator of equation (8) represents the integral in the in-channel (fuel rod) portion of the assembly, and the denominator of equation (8) represents the local output of the fuel rod area average that has undergone the combustion history. And shows the renormalization of the local output distribution. Since the local output distribution is a relative output obtained by normalizing the average value of the outputs of all the fuel rods in the fuel node to 1.0, the average value of p ^∞ (r) is 1.0.

In equation (8), assuming that the change in the heat group fission cross section is caused by the spectral history within the fuel node and the burnup distribution within the fuel node, the combustion history correction δp (r) for local power peaking for the candidate is It can be expressed by the following formula.

[0109]

[Equation 25] The meaning of the integration of the first term on the right side is the same as in equation (5), and represents the deviation from 1 of the average spectral history of the fuel rod portion of the assembly.

The first term in [] on the right side is that the deviation from the single-assembly combustion calculation of the local output peaking due to the deviation from 1 of the spectrum history is different from the deviation of the spectrum history from 1 in the fuel rod r. Is proportional to the difference in deviation from 1 in the spectral history of the fuel rod partial average. The local output peaking is a relative output in which the average value of all fuel rods in the fuel node is normalized to 1.0. Therefore, if the spectrum history in the fuel node is uniform, the effect of the spectrum history is for each fuel rod. , Which means that the local output does not change due to this.

As a candidate, it is appropriate to divide the assembly into several regions in the single-assembly combustion calculation and use the fuel rods having the maximum local power peaking in each region. In this way, by correcting the candidates, the time required for the fuel rod output calculation can be significantly shortened. This is particularly important when applying the present invention to online core performance monitoring.

Still another modification of the fuel node local power distribution calculation device 16 will be described.

Local output distribution apparatus 1 shown in this modification
No. 6 uses the fuel rod output that takes into account the change in the nuclear constant in the fuel node due to the combustion due to the neutron exchange (spectral interference effect) with the adjacent fuel node, and burns the fuel rod (burn). It is to calculate the critical output ratio, which is a monitoring index for (out). Here, the limit power ratio is defined as the ratio of the assembly output (limit output) at which the fuel rod undergoes a boiling transition and the current assembly output. The limit output ratio is, for example, “Toshiba Corporation: Outline of Boiling Water Nuclear Power Plant GETAB, TLR-009 (Revision 1), 1
As described in "June 981", is a function of the core pressure, boiling length, cooling water mass flux, heating length, thermal equivalent diameter, and the R-factor that characterizes the local power distribution in the fuel assembly. It is calculated using the "marginal quality correlation equation".

In the present embodiment, the fuel rod output in consideration of the change in the nuclear constant in the fuel node due to the combustion due to the exchange of neutrons (spectral interference effect) with the adjacent fuel node, By calculating the factors, it becomes possible to improve the calculation accuracy of the limiting output ratio.

In yet another modification of the local power distribution calculating device 16 in the fuel node, as will be described later, the change of the nuclear constant in the fuel node due to the combustion of the control rod inserted in the assembly is considered. By calculating the R factor using the above-mentioned fuel rod output, it is possible to accurately calculate the limit output ratio, which is a fuel rod burnout monitoring index.

Next, another embodiment of the reactor core performance calculation apparatus according to the present invention will be described.

FIG. 9 is a block diagram of another embodiment of the reactor core performance calculation apparatus according to the present invention.

Core performance calculation apparatus 1 shown in this embodiment
OA is the core performance calculation device 10 shown in FIG. 1 provided with a control rod history calculation device 25 in the fuel node and a nuclear constant control rod history correction calculation device 26 in the fuel node.

Core performance calculator 10A shown in FIG.
In addition to the core performance calculation device 10 shown in FIG. 1, a control rod history calculation device 25 in the fuel node is provided, and this control rod history calculation device 25 integrates the period in which control rods are inserted and the period after withdrawal. The calculation result is input to the fuel node core constant control rod history correction calculation device 26. The control rod history correction device 26 takes in the control rod history from the control rod history calculation device 25 and calculates the correction amount for the nuclear constant of the fuel node. These correction results are fed back to the core three-dimensional neutron diffusion calculation device 11, and the neutron flux distribution in the core 10 is calculated by convergence calculation.

In addition, the fuel node local power distribution calculation device 16 provided in the core performance calculation device 10A uses the spectrum history correction calculation device 15 to calculate the nuclear constant correction amount based on the spectrum history in the fuel node and the burnup distribution. The nuclear constant control rod history correction amount is input from the nuclear constant control rod history correction device 26 in the fuel node, and the local output distribution in the fuel node is calculated.

Further, the control rod history calculating device 25 in the fuel node provided in the core performance calculating device 10A integrates the period after inserting the control rod and the period after withdrawing the control rod to calculate the control rod history. The control rod calculator 26 is related to the calculation of the control rod history.

The hysteresis effect associated with the burning of the control rods inserted in the assembly is a kind of spectral hysteresis effect, but the spatial localization of the control rods causes the plutonium in the periphery of the control rods. There is a large delay in the accumulation and combustion of uranium. An example of this is shown in FIG.

In FIG. 10, a black circle indicates an infinite multiplication factor when a control rod is inserted and burned while the burnup of a certain assembly is from 0 GWd / st to 10 GWd / st, and then the control rod is pulled out and burned. And the difference in infinite multiplication factor when burning without a control rod from the beginning. The hysteresis void ratio (VH) is 40% in both cases. Immediately after the control rod is pulled out, the change in infinite multiplication factor jumps because the control rod is pulled out, and the importance around the control rod, where plutonium was large, increased.

[0124] On the other hand, the triangle marks indicate an example in which the spectrum history sensitivity coefficient extracted from void history calculation is used to calculate the spectrum history of the aggregate average when the control rod is inserted. Although the control rod hysteresis effect is reproduced fairly well during the control rod insertion period for the same aggregate average spectral hysteresis change amount, it can be seen that the error after pulling out the control rod is large. This is because the spatial distribution of the spectrum history due to the insertion of the control rod is not taken into consideration as described above.

Therefore, in this control rod history calculation device, in order to correctly evaluate the spatial distribution of the spectrum history due to the control rod insertion, it is given by non-homogeneous detailed calculation. Using the distribution of the ratio of the thermal neutron flux to the fast neutron flux in the assembly when the control rod is inserted, the control rod history effect is calculated by calculating the spectrum history in the control rod insertion fuel node. Is calculated using. Thereby, the control rod history effect can be treated as a spectrum history effect in a unified manner.

Further, the nuclear constant control rod history calculation device 25 in the fuel node may be constructed as follows.

In the embodiment of the control rod history calculation device 25, the control rod history is directly corrected without using the spectrum history sensitivity coefficient in consideration of the peculiarity of the control rod history effect, and the control rod history calculation is performed. The spectrum interference history is corrected by the interaction of neutrons (spectral interference effect). Regarding the control rod history effect in the example, the control rod history effect is calculated by a function of the period during which the control rod is inserted and burned, and the combustion period after the control rod comes off, and the neutron leaks from the adjacent node. The spectral history effect due to is calculated using the spectral history sensitivity coefficient.

As shown in FIG. 10, the control rod hysteresis effect is calculated by a quadratic equation of burnup of the control rod during insertion of the control rod, and after pulling out the control rod, after pulling out the control rod. It can be expressed as an exponential function of burnup. As a result, it becomes possible to consider the locality of the control rod history effect and to make calculations consistent with the spectrum interference history correction.

On the other hand, the nuclear constant control rod history correction calculation device 26 in the fuel node receives the control rod history effect calculated by the control rod history calculation device 25 and directly corrects this control rod history effect.

This embodiment of the control rod history correction calculation device 26 considers the spatial distribution of the control rod history effect when directly correcting the control rod history effect. Figure 11
In the control rod history calculation of FIG. 10, shows the difference (%) between the local output distribution of the control rod immediately after pulling out the control rod and the local output distribution during normal combustion. From this, it can be seen that the local output distribution is the largest at the (1,1) fuel rod closest to the center of the cross control rod, and becomes smaller as it gets farther from the control rod. Therefore, the change in the nuclear constant due to the control rod hysteresis effect from the single aggregate combustion calculation is corrected as a function of the pre-calculated distance from the control rod.

As described above, the core performance calculation device is the spectrum distribution calculation device 13 and the spectrum history calculation device 1.
8, a burnup distribution calculation device 19, a spectrum history correction calculation device 20, a burnup distribution correction calculation device 21, a control rod history calculation device 25, and a local power distribution calculation device 16 are provided, and the effective multiplication factor of the core by core performance calculation is provided. And a comparison of the effect of combustion history on the maximum linear power density and that by the whole core detailed calculation is shown in 12. This core performance calculation device 10A is
The detailed calculation is reproduced well.

[0132]

As described above, according to the method and apparatus for calculating core performance of a nuclear reactor according to the present invention, (1) the spatial distribution of the spectrum in the fuel node is corrected to a known value by the above-mentioned configuration. It is possible to reduce the calculation time by making a decision only in the vicinity of the target node using the one-group model, (2)
The calculation accuracy of the spectrum history effect can be improved by performing the non-separable expansion of the spectrum distribution and the spectrum history distribution in the node of interest by the analytical solution of the diffusion equation. (3) Combustion in the fuel node The degree distribution is
The accuracy of the burnup distribution effect can be improved by separating the contribution of the slope of the high-speed group and the contribution of the distortion of the spectrum, and (4) the sensitivity coefficient for the burnup distribution and the spectrum history distribution of the nuclear constants can be improved. The calculation accuracy can be improved, and (5) the effect of the spectrum interference history and the effect of the control rod history can be combined.

This makes it possible to accurately calculate the power distribution (neutron flux distribution) in the core and the fuel rod power distribution in the fuel node.

In the reactor core performance calculation apparatus according to the first aspect, even when the assembly burns due to neutron exchange with another assembly having a different spectrum in the core, a single assembly is produced. Correct the combustion history effect using only the nuclear constants given by body combustion calculation, and in a sufficiently short calculation time,
It is possible to accurately calculate the power distribution in the core and the fuel rod power distribution in the assembly.

In the reactor core performance calculation device according to the second aspect, the spectral distribution and burnup distribution in the fuel node are obtained using an integrating device, and the spectral distribution and burnup distribution are corrected by the correction calculation device. Therefore, the spectral distribution and burnup distribution in the fuel node can be accurately obtained, and the power distribution in the core and the fuel rod power distribution in the fuel node can be quickly and accurately obtained.

According to the reactor core performance calculation apparatus of the third aspect, the assembly burns by receiving neutron exchange with another assembly having a different spectrum in the core,
Even if a control rod is inserted into the assembly and burns, the combustion history effect is corrected using only the nuclear constants given by the single assembly combustion calculation, and the power distribution in the core is accurately calculated in a sufficiently short calculation time. And the fuel rod power distribution in the assembly can be calculated.

In the method of calculating core performance of a reactor according to claim 4, the neutron spectrum at the center of the fuel node (small volume) of the assembly size utilizes the property of approaching the asymptotic value in the single assembly combustion calculation. Then, the neutron spectrum in the node of interest can be calculated by solving the diffusion equation in a small region in the core by the spectrum distribution calculation device, which enables the fuel node to be calculated without performing the total core calculation for thermal neutrons. The spectrum history inside can be obtained accurately, and the calculation time can be shortened significantly.

In the core performance calculation method for a nuclear reactor according to claim 5, the spectrum distribution calculation device is used to calculate the spectrum distribution in the fuel node (small volume) by the neutron spectrum at a point on the side of the fuel node. By analytically solving the diffusion equation with the value of as the boundary value and obtaining it by the non-decomposition type expansion, it is possible to further reduce the time for solving the diffusion equation and calculate the spectrum history.

In the core performance calculation method for a nuclear reactor according to claim 6, the expansion formula for the spectrum distribution in the fuel node (small volume) is applied to the spectrum history distribution, and the spectrum history calculation device is used to The spectrum history is obtained by the non-separable expansion based on the values on the sides, and the accuracy of the spectrum history distribution is improved.

In the reactor core performance calculation method according to claim 7, the contribution of the inclination of the high speed group and the contribution of spectrum interference to the burnup distribution in the fuel node due to the exchange of neutrons with the adjacent node are separated. By doing so, the burnup distribution in the fuel node can be accurately calculated by the burnup distribution calculation device, and the correction accuracy of the nuclear constant due to the burnup distribution in the fuel node can be improved. is there.

In the core performance calculation method for a nuclear reactor according to claim 8, the sensitivity coefficient for the spectrum history of the nuclear constant used for correcting the spectrum interference history due to the exchange of neutrons with the adjacent node The sensitivity coefficient is improved by expressing the sensitivity coefficient as a function of the burnup and the history water density or the history void ratio by the spectral history correction calculation device in consideration of the dependency of the above.

In the core performance calculation method for a nuclear reactor according to claim 9, the sensitivity coefficient is obtained by an aggregate combustion calculation in consideration of neutron leakage or leakage by a spectrum history correction calculation device (neutrons with adjacent nodes). It is used to correct the spectrum interference history due to the exchange of the), and the accuracy of the sensitivity coefficient for the spectrum history of the nuclear constant is improved.

In the method of calculating core performance of a nuclear reactor according to claim 10, the burnup dependency correction of the nuclear constant of the burnup distribution in the fuel node due to the exchange of neutrons with the adjacent leads is corrected. Calculated by unit cell combustion calculation using a calculation device, or by polynomial approximation (quadratic fit) regarding burnup from burnup dependency at the point where burnup progressed with respect to the aggregate constant The accuracy of the burnup dependency of the nuclear constant used to correct the change of is improved.

In the core performance calculation method for a nuclear reactor according to claim 11, the sensitivity coefficient for the burnup of the nuclear constant used for correction and the burnup average value is compared with that of an ordinary fuel rod by using a burnup distribution correction calculation device. It is given separately with fuel rods containing burnable poisons to improve the accuracy of the nuclear constant used for correction by the spectrum history and burnup distribution in the fuel node due to the exchange of neutrons with adjacent nodes. .

In the core performance calculation method for a nuclear reactor according to claim 12, when the control rod is inserted, which is given by the non-homogeneous detailed calculation, in order to correctly evaluate the spatial distribution of the spectrum history due to the control rod insertion. By using the distribution of the ratio of the thermal neutron flux to the fast neutron flux in the assembly of, the control rod history calculation device calculates the spectrum history in the control rod insertion node to determine the control rod history effect as the spectrum history sensitivity coefficient. The control rod hysteresis effect can be treated as a spectrum hysteresis effect in a unified manner.

In the reactor core performance calculation method according to the thirteenth aspect, in consideration of the locality of the control rod history effect, the control rod history calculation device uses the spectrum history coefficient to calculate the control rod history. It is possible to calculate without any contradiction by directly correcting the spectrum interference history by correcting the spectrum interference history by exchanging neutrons with the adjacent aggregate.

In the reactor core performance calculation method according to the fourteenth aspect, the spatial distribution of the control rod hysteresis effect is taken into consideration, and the control rod hysteresis correction calculation device is used to calculate the nuclear constant based on the control rod hysteresis effect. The change from the one-bundle combustion calculation is calculated in advance and corrected as a function of the distance from the control rod,
The control rod history effect is directly corrected.

In the reactor core performance calculation method according to the fifteenth aspect, when correcting the change in the fuel rod output in the fuel node, the correction for the local output of the fuel rod is performed by using the local output distribution calculation device. The local output distribution can be directly corrected easily without using the sensitivity coefficient for the burnup and spectral history of the local output of the rod and for the nuclear constants such as the thermal group nuclear separation cross section. is there.

In the core performance calculation method for a nuclear reactor according to claim 16, when correcting a change in fuel rod output in a node due to burning due to exchange of neutrons with an adjacent node, a local Using the power distribution calculator, select a few fuel rods that are predicted to be the most thermally demanding, and consider the change from the single-assembly calculation of nuclear constants for this candidate. By calculating the local power peaking described above, the time required to calculate the fuel rod power can be significantly shortened. This can be suitably applied to online core performance monitoring.

In the method for calculating core performance of a nuclear reactor according to claim 17, the fuel rod output in consideration of the change in the nuclear constant in the fuel node due to the combustion due to the exchange of neutrons with the adjacent node. The calculation accuracy of the limit output ratio can be improved by calculating the limit output ratio, which is a monitoring index of burnout of the fuel rods, by using the local output distribution calculation device.

In the method for calculating core performance of a nuclear reactor according to claim 18, the fuel rod output is used in consideration of the change in the nuclear constant in the fuel node due to the combustion of the control rod inserted in the assembly. Then, by calculating the R factor by the local power distribution calculation device, it is possible to accurately calculate the limit power ratio, which is a fuel rod burnout monitoring index.

As described above, according to the core performance calculation method and apparatus of the present invention, (1) the spectral history distribution in the fuel node is used only in the vicinity of the node of interest, and the non-separable expansion by the analytical solution of the diffusion equation is used. Therefore, the accuracy of the spectrum history effect is improved.

(2) Regarding the burnup distribution effect, the burnup distribution within the fuel node can be calculated separately for the contribution of the inclination of the high speed group and the contribution of the spectral interference, and the calculation accuracy of the burnup distribution within the fuel node is improved. .

(3) Regarding the control rod history effect, the spectrum history effect is calculated in consideration of the locality of the effect.

Therefore, it is possible to correct the spectrum interference history effect, burnup distribution effect, and control rod history effect with respect to the power distribution in the core and the fuel rod power distribution in the fuel node accurately in a short time.

[Brief description of drawings]

FIG. 1 is a block diagram showing an embodiment of a reactor core performance calculation apparatus according to the present invention.

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

FIG. 3 is a diagram showing a spectrum change of a neutron spectrum in a fuel node from a single-aggregate combustion calculation device.

FIG. 4 shows an example of a change in fuel rod output within a fuel node from a single aggregate, where (a) is a difference from a single aggregate combustion calculation of local power peaking, and (b) is a neutron spectrum. Strain contribution, (c) is a diagram showing the contribution of the inclination of the fast neutron flux.

FIG. 5 is a diagram showing a sensitivity coefficient (infinite multiplication factor) with respect to a spectrum history of a nuclear constant in a fuel node.

FIG. 6 is a diagram showing a sensitivity coefficient (heat group fission cross section) with respect to a spectrum history of a nuclear constant in a fuel node.

FIG. 7 is a diagram showing a comparison effect of core eigenvalues when a hysteresis history density dependence is given to a spectrum hysteresis sensitivity coefficient.

FIG. 8 is a diagram showing burnup dependence of a heat group nuclear fission cross section.

FIG. 9 is a block diagram showing another embodiment of the reactor core performance calculation device according to the present invention.

FIG. 10 is a view showing a comparison of control rod history effects with respect to infinite multiplication factors.

FIG. 11: Local output of fuel rod (local output peaking coefficient)
Showing the control rod history effect for the.

FIG. 12 is a diagram showing the spectrum interference history effect by comparing the accuracy of the reactor core performance calculation apparatus according to the present invention with the inhomogeneous detailed calculation of the reactor core.

[Explanation of symbols]

10,10A Reactor core performance calculator 11 core 12 Core 3D Neutron Diffusion Calculator 13 Spectral distribution calculator in fuel node 14 Accumulator 15 Correction calculator 16 Local power distribution calculator in fuel node 18 Intrafuel node spectrum history calculator 19 Burnup distribution calculator in fuel node 20 Nuclear constant spectrum history correction calculation device in fuel node 21 Nuclear constant burnup distribution history correction calculator for fuel node 23,24 Fuel Node 25 Control rod history calculator in fuel node 26 Fuel Node Core Constant Control Rod History Correction Calculation Device

Continuation of the front page (56) Reference JP-A-5-80182 (JP, A) JP-A-4-236397 (JP, A) JP-A-3-264896 (JP, A) JP-A-4-142498 (JP , A) JP 62-184392 (JP, A) Takeichi Toshikazu, New method for calculating power distribution in light water reactor, Nuclear Industry, Japan, June 1987, No. 33 Vol. 6, P58- 64 Etsuro Saji Tatsuya Iwamoto Hiromi Maruyama Yoshitoshi Tahara Masaaki Mori, Advancement of Commercial Light Water Reactor Nuclear Calculation Methods, Journal of the Atomic Energy Society of Japan, Japan, June 1994, No. 36 Vol. 6, P484-494 T.I. Iwamoto, M .; Yamamoto, M .; Tsukii, Verification of LOGOS N oder Methods with Heterogeneous Burnu p Calculations for a BWR Core, Tran. 71, P251-253 (58) Fields investigated (Int.Cl. ⁷ , DB name) G21C 17/00

Claims (16)

(57) [Claims] 1. A core 3 for calculating a nuclear constant in a fuel node by taking in core state parameters from the core to calculate a fast neutron flux distribution in the core and an asymptotic spectrum of the fuel node.
-Dimensional neutron diffusion calculation device, spectrum distribution calculation device for obtaining the spectrum distribution in the fuel node by inputting fast neutron flux distribution and asymptotic spectrum from this core three-dimensional neutron diffusion calculation device, and spectrum from this spectrum distribution calculation device Average of the ratio of the distribution and the asymptotic spectrum of the spectral distribution in the fuel node to the burnup E using the asymptotic spectrum, that is, And an integrator for calculating the burn-up distribution and the spectrum history SH (r), and the adjoining from the nuclear constant of the single-aggregate combustion calculation in the fuel node using the spectrum history and burn-up distribution from this integrator Spectral history correction amount to correct the change of the nuclear constant due to neutron exchange with the fuel node and the change of the nuclear constant due to the non-uniform combustion from the nuclear constant of the single-aggregate combustion calculation in the fuel node A correction calculation device for calculating a burnup distribution correction amount, and a local output distribution calculation device for inputting the spectrum history correction amount and the burnup distribution correction amount to obtain a correction for the local output in the fuel node and calculating a local output distribution And the core three-dimensional neutron diffusion calculation device feeds back the spectrum history correction amount and the burnup distribution correction amount from the correction calculation device to perform convergence calculation. Core performance calculation apparatus of the reactor, characterized in that to calculate the neutron flux distribution in the heart. 2. The integrating device inputs a spectral distribution in a fuel node and the asymptotic spectrum to calculate a spectrum history calculating device for calculating a spectrum history in the fuel node and a burnup distribution calculating device for calculating a burnup distribution. On the other hand, the correction calculation device includes a spectrum history calculation correction device that calculates a spectrum history correction amount for a fuel node nuclear constant and a burnup distribution correction calculation device that calculates a burnup distribution correction amount based on the output from the integration device. The reactor core performance calculation device according to claim 1, further comprising: 3. A core 3 for calculating a nuclear constant in a fuel node by taking in core state parameters from the core to calculate a fast neutron flux distribution in the core and an asymptotic spectrum of the fuel node.
-Dimensional neutron diffusion calculation device, spectrum distribution calculation device that obtains the spectrum distribution in the fuel node by inputting fast neutron flux distribution and asymptotic spectrum from this core three-dimensional neutron diffusion calculation device, and spectrum from this spectrum distribution calculation device Using the distribution, the average value of the ratio of the spectral distribution in the fuel node to the asymptotic spectrum for the burnup E, that is, And an integrator for calculating the burn-up distribution and the spectrum history SH (r), and the adjoining from the nuclear constant of the single-aggregate combustion calculation in the fuel node using the spectrum history and burn-up distribution from this integrator Spectral history correction amount to correct the change of the nuclear constant due to neutron exchange with the fuel node and the change of the nuclear constant due to the non-uniform combustion from the nuclear constant of the single-aggregate combustion calculation in the fuel node A correction calculation device that calculates the burnup distribution correction amount, and a control within the fuel node that is the spectrum history within the fuel node obtained from the spectral distribution within the fuel node during the control rod insertion period and its withdrawal period A control rod history calculation device that calculates rod history and a control rod history from this control rod history calculation device are input to set the correction amount for the nuclear constant of the fuel node. A control rod history correction calculation device for calculating, a spectrum history correction amount and a burnup distribution correction amount from the correction calculation device, and a control rod history correction amount from the control rod history correction calculation device are input to locally output in the fuel node. And a local power distribution calculation device for calculating a local power distribution by obtaining a correction to the core, and the core three-dimensional neutron diffusion calculation device includes a spectrum history correction amount and a burnup distribution correction amount and a control rod history correction from the correction calculation device. An apparatus for calculating core performance of a nuclear reactor, characterized in that a neutron flux distribution in a core is calculated by a convergence calculation by feeding back a control rod history correction amount from a calculator. 4. The neutron flux in the core is divided by dividing the core of the nuclear reactor into small volumes, and solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by the single-assembly combustion calculation. In the core performance calculation method of a nuclear reactor for calculating distribution and fuel rod power distribution in an assembly, a spectral distribution which is a distribution of a ratio of thermal neutron flux to fast neutron flux in the small volume, the small volume and surrounding it Obtained by solving a diffusion equation for a small region in the core composed of adjacent regions, and using this spectral distribution, the average value of the ratio of the small volume spectral distribution to the asymptotic spectrum with respect to burnup E, that is, From the single-aggregate combustion calculation of the nuclear constant due to the fact that the small volume burns in response to the exchange of neutrons with the adjacent small volume using this spectrum history SH (r) A method for calculating core performance of a nuclear reactor, which is characterized in that 5. An analytical solution obtained by analytically solving a neutron diffusion equation for a distribution of a ratio of a thermal neutron flux to a fast neutron flux in a small volume with a value of the ratio on the side of the small volume as a boundary value. The core performance calculation method for a nuclear reactor according to claim 4, which is obtained by a non-separable type expansion according to. 6. The neutron flux in the core is divided by dividing the core of the nuclear reactor into small volumes, and solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by the single-assembly combustion calculation. In the core performance calculation method of a nuclear reactor for calculating distribution and fuel rod power distribution in an assembly, a spectral distribution which is a distribution of a ratio of thermal neutron flux to fast neutron flux in the small volume, the small volume and surrounding it It is obtained by solving a diffusion equation for a small region in the core composed of adjacent regions, and using this spectral distribution, the ratio of the ratio of the small volume spectral distribution to the asymptotic spectrum to the burnup E, that is, The spectrum history SH (r) is obtained by non-separable expansion by an analytical solution of the neutron diffusion equation whose boundary value is a value on the side of the small volume, and the small volume is adjacent to the small volume by using this spectrum history. A method for calculating core performance of a nuclear reactor, characterized in that a change from the single-aggregate combustion calculation of the nuclear constant due to burning due to neutron interaction with a small volume is calculated. 7. The neutron flux in the core is divided by dividing the core of a nuclear reactor into small volumes, and solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by the single-assembly combustion calculation. In a core performance calculation method of a nuclear reactor for calculating distribution and fuel rod power distribution in an assembly, a single set of the nuclear constants resulting from the small volume burned by receiving neutron interaction with an adjacent small volume Obtain the change from the body combustion calculation, the distribution of burnup in the small volume based on the interaction of neutrons with the adjacent small volume, divided into components based on the slope of the fast neutron flux and components based on the distortion of the spectrum, high speed The component based on the slope of the neutron flux is obtained by a bipolynomial expansion whose boundary value is the value on the side of the small volume, and the component based on the distortion of the spectrum on the side of the small volume. Core performance calculation method of the reactor, characterized in that determined by the non-separation deployments by analytical solution of the neutron diffusion equation for the boundary value. 8. The neutron flux in the core is divided by dividing the core of the nuclear reactor into small volumes, and solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by the single-assembly combustion calculation. In the core performance calculation method of a reactor for calculating distribution and fuel rod power distribution in an assembly, the ratio of thermal neutron flux to fast neutron flux in the small volume
The spectral distribution , which is a distribution, is obtained by solving the diffusion equation for the small region in the core consisting of the small volume and the adjacent region surrounding it, and using this spectral distribution , the ratio of the ratio of the small volume spectral distribution to the asymptotic spectrum is calculated. Average value for burnup E, that is, Is obtained, the change from the single-aggregate combustion calculation of the nuclear constant due to the small volume burned by receiving neutron interaction with the adjacent small volume, the spectrum within the small volume A method for calculating core performance of a nuclear reactor, wherein the sensitivity coefficient for the spectral history is obtained by using a history and is a function of an average value for the burnup of the small volume and the burnup of the water density. 9. The neutron flux in the reactor core is divided into small volumes, and the neutron diffusion equation or the transport equation is solved by using the nuclear constant for the small volume given by the single assembly combustion calculation. In the reactor core performance calculation method of calculating the distribution and fuel rod power distribution in the assembly, a spectrum that is the ratio of thermal neutron flux to fast neutron flux in the small volume, from the small volume and the adjacent region surrounding it Is obtained by solving a diffusion equation for a small region in the core, and using this spectrum, the average value of the ratio of the small volume spectrum to the asymptotic spectrum with respect to burnup E, that is, Is obtained, the change from the single-aggregate combustion calculation of the nuclear constant due to the small volume burned by receiving neutron interaction with the adjacent small volume, the spectrum within the small volume A method for calculating core performance of a nuclear reactor, wherein a sensitivity coefficient for the spectral history is obtained by using history, and the sensitivity coefficient for the spectrum history is obtained by aggregate calculation in consideration of leakage or leakage of neutrons into the small volume. 10. The neutron flux in the core is divided by dividing the core of a nuclear reactor into small volumes, and using a nuclear constant for the small volume given by a single-assembly combustion calculation to solve a neutron diffusion equation or a transport equation. In a core performance calculation method of a nuclear reactor for calculating distribution and fuel rod power distribution in an assembly, a single set of the nuclear constants resulting from the small volume burned by receiving neutron interaction with an adjacent small volume A change from the body combustion calculation is obtained, and the sensitivity coefficient to the burnup of the nuclear constant used when obtaining this change is obtained by a unit cell combustion calculation consisting of only one fuel rod, or
A method for calculating core performance of a nuclear reactor, which is characterized by a polynomial approximation of burnup from the burnup dependence of the point at which burnup progresses with respect to the aggregate mean nuclear constant. 11. A neutron flux in a core is divided by dividing a core of a nuclear reactor into small volumes and solving a neutron diffusion equation or a transport equation using a nuclear constant for the small volume given by a single-assembly combustion calculation. In a core performance calculation method of a nuclear reactor for calculating distribution and fuel rod power distribution in an assembly, a single set of the nuclear constants resulting from the small volume burned by receiving neutron interaction with an adjacent small volume Obtaining the change from body combustion calculation, the sensitivity coefficient for the burnup of the nuclear constant used to determine this change and the burnup average value of the ratio of thermal neutron flux to fast neutron flux is the fuel rod that does not contain burnable poisons. A method for calculating core performance of a nuclear reactor, characterized in that fuel rods containing burnable poisons are given separately. 12. The neutron flux in the core is divided by dividing the core of the nuclear reactor into small volumes, and solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by the single assembly combustion calculation. In the core performance calculation method of the reactor for calculating the distribution and the fuel rod power distribution in the assembly, the change from the single assembly combustion calculation of the nuclear constant due to the combustion of the control rod inserted in the small volume is described. , The spectral distribution is the distribution of the ratio of the thermal neutron flux to the fast neutron flux in the small volume given by the aggregate inhomogeneous calculation when the control rod is inserted in the small volume, using this spectral distribution The mean value of the ratio of the small volume spectrum to the asymptotic spectrum to the burnup E, that is, A method for calculating core performance of a nuclear reactor characterized in that the spectrum history SH (r) is calculated and corrected using this spectrum history. 13. The amount of change in the nuclear constant in the small volume from the single aggregate combustion calculation due to the insertion of the control rod in the small volume is corrected by a correction coefficient according to the distance from the control rod. The reactor core performance calculation method according to claim 12, which is obtained by using the method. 14. A plurality of thermally strict fuel rods are selected as candidates in the small region of the core, and a change from a single-assembly combustion calculation of a nuclear constant is considered for these candidates. 14. The method for calculating core performance of a nuclear reactor according to claim 4, wherein the fuel rod output is calculated. 15. The neutron flux in the core is divided by dividing the core of a nuclear reactor into small volumes and solving the neutron diffusion equation or the transport equation using the nuclear constant for the small volume given by the single assembly combustion calculation. In the core performance calculation method of a nuclear reactor for calculating distribution and fuel rod power distribution in an assembly, a spectral distribution which is a distribution of a ratio of thermal neutron flux to fast neutron flux in the small volume, the small volume and surrounding it Obtained by solving a diffusion equation for a small region in the core composed of adjacent regions, and using this spectral distribution, the average value of the ratio of the small volume spectral distribution to the asymptotic spectrum to the burnup E, that is, From the single-mass combustion calculation of the nuclear constant due to the fact that the small volume burns in response to neutron interaction with an adjacent small volume using this spectrum history SH (r) The method for calculating the core performance of a reactor is characterized in that the critical power ratio, which is a fuel rod burnout monitoring index, is calculated using the fuel rod power that takes into account this change. 16. A control rod is added to the small volume in addition to the change from the single-aggregate combustion calculation of the nuclear constant due to the fact that the small volume burns in response to neutron exchange with an adjacent small volume. It is characterized in that a limit power ratio, which is a monitoring index of burnout of a fuel rod, is calculated by using a fuel rod output in consideration of a change from the single assembly combustion calculation of the nuclear constant caused by being inserted and burned. The core performance calculation method for a nuclear reactor according to claim 15.