JPH0875891A - Reactor core performance computing method and apparatus therefor - Google Patents

Abstract

PURPOSE: To compute output distribution inside a core and burnup output distribution inside a node highly precisely within a short time by a method to compute neutron flux distribution by carrying out feedback of both corrected results of spectrum history and specific burnup distribution. CONSTITUTION: A computing apparatus 13 for spectrum in a fuel node computes spectrum distribution analytically and sends the result to an integrating apparatus 14, and a computer 20 for correction of spectrum history for nuclear constant in fuel node and a computer 21 for correction of burnup distribution for nuclear constant in fuel node compute the spectrum history corrected result and the burnup distribution corrected result to the fuel node nuclear constant by receiving the spectrum history and the burnup distribution from the integrating apparatus 14, the results are sent back to a three-dimensional nuclear diffusion computer 12, and the neutron flux distribution in the core is computed by converging calculation. A computer 16 for calculation of local output distribution in fuel node receives the spectrum history corrected result and the burnup distribution corrected result to the nuclear constant in a fuel node from the spectrum history correcting apparatus 20 and computes the local output distribution in the fuel node.

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) of the ratio of the neutron spectrum of the fuel node to the asymptotic spectrum is added as a parameter, and the burnup average value (= asymptotic spectrum). / Fuel node neutron spectrum)
The homogenization nuclear constant is changed with respect to the deviation of. For example, the change δΣ _f2 (r) in the thermal group nuclear fission cross section at the position r in the fuel node is given by the following equation.

[0008]

[Equation 1] The spectrum history SH is an average value of the ratio of the 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 2]

[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. A three-dimensional neutron diffusion calculation device for the core that calculates the nuclear constants in the node and calculates the fast neutron flux distribution in the core and the asymptotic spectrum of the fuel node, and the fast neutron flux distribution and asymptotic spectrum from this neutron diffusion calculation device are input. And a spectrum distribution calculation device for calculating the spectrum distribution in the fuel node, an integration device that calculates the spectrum history and burnup distribution in the fuel node using the spectrum distribution from this spectrum distribution calculation device, Correction to calculate spectral history correction amount and burnup distribution correction amount for nuclear constants in fuel node using spectrum distribution and burnup distribution And calculation device, and a local power distribution calculation unit that calculates the local power distribution of the spectrum history correction amount and the burn distribution correction amount by inputting the fuel node,
The core three-dimensional neutron diffusion calculation device is configured to feed back the spectrum history correction amount and the burnup distribution correction amount from the correction calculation device to calculate the neutron flux distribution in the core by convergence calculation.

In order to solve the above-mentioned problems, in the reactor core performance calculation apparatus according to the present invention, as described in claim 2, the integrator inputs the spectral distribution in the fuel node and The correction history calculation device includes a spectrum history calculation device that calculates the spectrum history in the node and a burnup distribution calculation device that calculates the burnup distribution, while the correction calculation device uses the output from the integration device to correct the spectrum history correction amount for the fuel node nuclear constant. 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. Number, calculating the fast neutron flux distribution in the core and the asymptotic spectrum of the fuel node, and a three-dimensional neutron diffusion calculation device of the core, and the fast neutron flux distribution and asymptotic spectrum from this neutron diffusion calculation device are input to the fuel node , A spectrum distribution calculation device for obtaining the spectrum distribution in the fuel node, an integration device for calculating the spectrum history and burnup distribution in the fuel node using the spectrum distribution from this spectrum distribution calculation device, and the spectrum distribution and combustion from this integration device And a correction calculation device for calculating the correction amount of the spectrum history and the correction amount of the burnup distribution for the core constant in the fuel node using A control rod history calculator that calculates the control rod history in the fuel node by integrating the control rod insertion period and its withdrawal period, and the nuclear constant of the fuel node by inputting the control rod history from this control rod history calculator Control rod history correction calculation device for calculating a correction amount for the fuel node, and a spectrum history correction amount and 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. And a local power distribution calculation device for calculating a local power distribution within the core, wherein the core three-dimensional neutron diffusion calculation device includes a spectrum history correction amount and a burnup distribution correction amount from the correction calculation device, and a control rod history correction calculation device. The neutron flux distribution in the core is calculated by a convergence calculation by feeding back the control rod history correction amount.

On the other hand, in order to solve the above-mentioned problems, the method for calculating the core performance of a reactor according to the present invention divides the core of the reactor into small volumes (fuel nodes) as described in claim 4. , The nuclear constant for the small volume given by the single-assembly combustion calculation is used to solve the neutron diffusion equation or the transport equation to calculate the neutron flux distribution in the core and the fuel rod power distribution in the assembly. In the core performance calculation method, the change from the single-aggregate combustion calculation of the nuclear constant due to the small volume burning due to the interaction of neutrons with the adjacent small volume, fast neutrons in the small volume Obtained by using the burnup average value of the ratio of thermal neutron flux to the flux, the distribution of the ratio of thermal neutrons to fast neutrons in the small volume used for the correction, the small volume and the adjacent region surrounding it A method of obtaining by solving the diffusion equation for the small region Ranaru the core.

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 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, thermal neutrons for fast neutron flux in the small volume Obtained using the burnup average value of the ratio of the flux, the distribution of the burnup average value of the ratio of the thermal neutron flux to the fast neutron flux in the small volume, the burnup average value on the side of the small volume as a boundary value A method of obtaining a non-separable deployments by analytical solution of the neutron diffusion equations.

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 change from the single-aggregate combustion calculation of the nuclear constant due to the small volume burning due to the interaction of neutrons with the adjacent small volume, a fast neutron flux in the small volume. Is obtained by using the burnup average value of the ratio of thermal neutron flux to, the sensitivity coefficient to the burnup average value of the ratio of thermal neutron flux to the fast neutron flux of the nuclear constant of the small volume used for the correction, of the small volume It is a method that is a function of the burnup and the average burnup of 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, 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, thermal neutrons for fast neutron flux in the small volume Obtained using the burnup average value of the ratio of the flux, the sensitivity coefficient for the burnup average value of the ratio of the thermal neutron flux to the fast neutron flux of the nuclear constant of the small volume used for the correction, the small volume A method of obtaining the neutron leakage or aggregate calculation considering leakage for.

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 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, sensitivity to the burnup of the nuclear constant used for correction The coefficient is calculated by a unit cell combustion calculation consisting of only one fuel rod, or is related to the burnup from the burnup dependence of the point at which the burnup is advanced with respect to the aggregate mean nuclear constant. A method of obtaining by Polynomial approximation.

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. The sensitivity coefficient for the burnup of the nuclear constant and the burnup of the ratio of the thermal neutron flux to the fast neutron flux, which are used to correct the change from the aggregate combustion calculation, and the burnup average value are fuel rods and burnable poisons that do not contain burnable poisons. It is a method of separately providing to fuel rods including.

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. Is a method of correcting using the burnup average value of the ratio of the thermal neutron flux to the fast neutron flux calculated using the distribution of the ratio of the thermal neutron flux to the fast neutron flux in the small volume given by .

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. 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, for 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, the fast neutron flux in the small volume Is corrected by using the burnup average value of the ratio of thermal neutron flux to, and the change from the single-aggregate combustion calculation of the nuclear constant caused by the combustion of the control rod inserted in the small volume is considered. Te is a method of adding a correction using the burnup accumulated value after burn accumulated value and the control rod of the period in which the control rod is inserted is pulled out.

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 14, characterized in that the control rod is inserted into the small volume. 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.

Further, 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 15. Using the average nuclear constant for the small volume given by body burning calculation, in the reactor core performance calculation method of calculating the fuel rod power distribution in the assembly by solving the neutron diffusion equation or the transport equation, the small volume is The fuel rod output in the small volume due to the burning due to the interaction of neutrons with the adjacent small volume is corrected, and the local output of the fuel rod is corrected by the local peaking burnup and fast neutrons of each fuel rod. This is a method that uses the sensitivity coefficient for the burnup average value of the ratio of thermal neutron flux to flux.

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 16, has several fuel rods that are thermally severe in a small region of the core. Is selected as a candidate, and the fuel rod power is calculated for this candidate in consideration of the change from the single-assembly combustion calculation of the nuclear constant.

Further, 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 assembly as described in claim 17. 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, using the fuel rod output considering 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, This is a method of calculating a limit output ratio, which is a monitoring index for burnout.

Further, in order to solve the above-mentioned problems, the method for calculating the core performance of a nuclear reactor according to the present invention sets forth a claim 18
, 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 core performance calculation method of a nuclear reactor for calculating a bundle distribution and a fuel rod power distribution in an assembly, a single assembly combustion calculation of the nuclear constant resulting from the burning of a control rod inserted in the small volume is performed. This is a method of calculating the limiting power ratio, which is a monitoring index of burnout of the fuel rod, by using the fuel rod output in consideration of the change.

[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.

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 is used to calculate the control rod history by using the spectrum history coefficient. 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 core performance calculation method for a nuclear reactor 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 calculating 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 method of calculating core performance of a nuclear reactor according to claim 16, when correcting a change in fuel rod output in a node due to combustion 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 seventeenth 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 eighteenth 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 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 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]

(Equation 3) 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 width of the fuel rod, and at the center of the fuel node of the assembly size,
The spectrum utilizes 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 spectrum as a boundary condition at the center of each fuel node. As a result, without performing a full core calculation for thermal neutrons,
The spectrum history in the fuel node can be obtained with high accuracy, and the calculation time can be greatly reduced.

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 distribution within the fuel node. In the evaluation method of Lempe et al., The spectrum history inside the fuel node was developed one-dimensionally from the value of the spectrum history at the side of the fuel node, so the accuracy was insufficient. In this spectrum history calculation device 18, the expansion formula (3) for the spectrum distribution in the fuel node is also applied to the spectrum history distribution, and it is obtained by the non-separable expansion from the value on the side of the fuel node.

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

[Equation 4] d _{1 to 8} are history coefficients that can be determined by the spectrum history at points on the sides of the fuel node 24 indicated by white circles in FIG. The history coefficients d _{1 to 8} can improve the calculation accuracy for the two-dimensional distribution of the spectrum history in the fuel node. As an example showing the effect of this spectrum distribution calculation device 13, for the deviation from 1 of the spectrum history in the highly enriched fuel surrounded by the less enriched fuel,
FIG. 3 shows a comparison between the calculation example (black circle) and the detailed calculation (white circle) by the spectrum history calculation device 18.

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, and the burnup distribution within the fuel node is calculated as the average value of the burnup and the volume average of the nodes. 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 5) The deviation δE _SH (r) of the spectral 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 6) 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 7)

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 in which neutron leakage or leakage is considered.

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 corrects the local power of the fuel rods in a method of radiating the change in the fuel rod power in the fuel node due to the combustion caused by the exchange of neutrons with the adjacent fuel node. This is performed by using the sensitivity coefficient for the burnup of the local output of the fuel rod and the spectrum history. 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 fission cross section Σ _f2 of the heat group of the fuel rod r of _interest changes due to the combustion history effect is

[Equation 8] 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 of the heat group cross-sectional area is caused by the spectral history in the fuel node and the burnup distribution in the fuel node, the combustion history correction δp (r) of the local output peaking for the candidate is as follows. Can be expressed as

[0109]

[Equation 9] 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 the 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 a function of the R-factor characterizing core pressure, boiling length, cooling water mass flux, heating length, thermal equivalent diameter, and local power distribution within the fuel assembly as shown in June 981. 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, black circles represent infinite multiplication factors when a control rod is inserted and burned for a certain assembly while the burnup 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 constant 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 method of calculating core performance of a nuclear reactor according to claim 16, in the case of 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 three-dimensional neutron diffusion calculator 13 Fuel node spectral distribution calculator 14 Accumulator 15 Correction calculator 16 Fuel node local power distribution calculator 18 Fuel node spectrum history Calculator 19 Fuel node burnup distribution calculator 20 Fuel node core constant spectrum history correction calculator 21 Fuel node core constant burnup distribution history calculator 23, 24 Fuel node 25 Fuel rod control rod history calculator 26 Fuel node Inner core control rod history correction calculator

Claims (18)

[Claims] 1. A core three-dimensional neutron diffusion calculation device for calculating a nuclear constant in a fuel node by taking in core state parameters from the core, and calculating a fast neutron flux distribution in the core and an asymptotic spectrum of the fuel node. Spectral distribution calculator that obtains the spectral distribution in the fuel node by inputting the fast neutron flux distribution and asymptotic spectrum from the neutron diffusion calculator, and the spectral history in the fuel node using the spectral distribution from this spectral distribution calculator An integrator that calculates a burnup distribution, a correction calculator that calculates a spectrum history correction amount and a burnup distribution correction amount for a core constant in a fuel node using the spectrum distribution and the burnup distribution from the integration device, and the spectrum history Local output that calculates the local power distribution in the fuel node by inputting the correction amount and burnup distribution correction amount A distribution calculation device, and the core 3
The dimensional neutron diffusion calculation device is characterized in that the spectrum history correction amount and the burnup distribution correction amount from the correction calculation device are fed back to calculate the neutron flux distribution in the core by convergence calculation. Performance calculator. 2. The integrating device comprises a spectrum history calculating device for inputting a spectrum distribution in the fuel node to calculate a spectrum history in the fuel node and a burnup distribution calculating device for calculating a burnup distribution, The correction calculation device comprises: a spectrum history calculation correction device that calculates a spectrum history correction amount for the 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. 1. The reactor core performance calculation device according to 1. 3. A core three-dimensional neutron diffusion calculation device for calculating a nuclear constant in a fuel node by taking in core state parameters from the core, and calculating a fast neutron flux distribution in the core and an asymptotic spectrum of the fuel node. Spectral distribution calculator that obtains the spectral distribution in the fuel node by inputting the fast neutron flux distribution and asymptotic spectrum from the neutron diffusion calculator, and the spectral history in the fuel node using the spectral distribution from this spectral distribution calculator An integrating device for calculating the burnup distribution, a correction calculating device for calculating the spectrum history correction amount and the burnup distribution correction amount for the nuclear constants in the fuel node using the spectrum distribution and the burnup distribution from this integrating device, and the control rod A control rod history calculator for calculating the control rod history in the fuel node by integrating the insertion period and its withdrawal period; A control rod history correction calculation device for inputting the control rod history from this control rod history calculation device to calculate a correction amount for the nuclear constant of the fuel node, and a spectrum history correction amount and burnup distribution correction amount from the correction rod calculation device. And a local power distribution calculation device for calculating the local power distribution in the fuel node by inputting the control rod history correction amount from the control rod history correction calculation device, wherein the core three-dimensional neutron diffusion calculation device is the correction calculation device. It is characterized in that the neutron flux distribution in the core is calculated by convergence calculation by feeding back the correction amount of spectrum history and the correction amount of burnup distribution from the control rod and the correction amount of control rod history from the control rod history correction calculation device. Core performance calculator for nuclear reactors. 4. The core of a nuclear reactor is divided into small volumes (fuel nodes), and the neutron diffusion equation or transport equation is solved by using the nuclear constant for the small volume given by a single assembly combustion calculation. In the reactor core performance calculation method for calculating the neutron flux distribution in a fuel cell and the fuel rod power distribution in an assembly, the nuclear constant resulting from the fact that the small volume burns by receiving neutron exchange with an adjacent small volume The change from the single-aggregate combustion calculation of is obtained by using the burnup average value of the ratio of thermal neutron flux to fast neutron flux in the small volume, and thermal neutrons for fast neutrons in the small volume used for the correction Is calculated by solving a diffusion equation for a small region in the core consisting of the small volume and the adjacent region surrounding the small volume. . 5. The distribution of the ratio of thermal neutrons to fast neutrons in a small volume is determined by an analytical solution obtained by analytically solving a neutron diffusion equation with the value of the ratio on the side of the small volume as a boundary value. The method for calculating core performance of a nuclear reactor according to claim 4, wherein the core performance calculation method is obtained by separation type expansion. 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 single set of the nuclear constants resulting from the small volume burning due to interaction of neutrons with adjacent small volume Change from body combustion calculation, using the burnup average value of the ratio of thermal neutron flux to the fast neutron flux in the small volume, burnup average of the ratio of thermal neutron flux to the fast neutron flux in the small volume A method for calculating core performance of a nuclear reactor, characterized in that a distribution of values is obtained by a non-separable expansion based on an analytical solution of a neutron diffusion equation having a burnup average value on the side of the small volume as a boundary value. 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 the 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 burning due to interaction of neutrons with 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 bipolynomial expansion with the value on the side of the small volume as the boundary value, and the component based on the distortion of the spectrum is the value on the side of the small volume. A method for calculating core performance of a nuclear reactor characterized by the non-separable expansion based on the analytical solution of the neutron diffusion equation with a boundary value of. 8. 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 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 burning due to interaction of neutrons with adjacent small volume Change from body burning calculation, obtained by using the burnup average value of the ratio of thermal neutron flux to fast neutron flux in the small volume, thermal neutron flux for fast neutron flux of the nuclear constant of the small volume used for the correction The method for calculating core performance of a nuclear reactor, wherein the sensitivity coefficient of the ratio of the above to the average burnup of the small volume is a function of the average burnup of the small volume and the average 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 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 burning due to interaction of neutrons with adjacent small volume Change from body burning calculation, obtained by using the burnup average value of the ratio of thermal neutron flux to fast neutron flux in the small volume, thermal neutron flux for fast neutron flux of the nuclear constant of the small volume used for the correction Sensitivity coefficient for the burnup average value of the ratio of is determined by the aggregate calculation considering leakage or leakage of neutrons in the small volume. Noh calculation method. 10. 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-aggregate combustion calculation. In the 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 burning due to interaction of neutrons with adjacent small volume The sensitivity coefficient to the burnup of the nuclear constant used for correction is calculated from the change from the body burnup calculation, or is calculated by the unit cell burnup calculation consisting of only one fuel rod, or the burnup to the aggregate average nuclear constant is advanced. A method for calculating core performance of a nuclear reactor, which is characterized by polynomial approximation of burnup from the burnup dependency of points. 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 the 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 burning due to interaction of neutrons with adjacent small volume The sensitivity coefficient for the burnup of the nuclear constant used for correction and the burnup of the ratio of the thermal neutron flux to the fast neutron flux to the average burnup of the burnup used for correction is calculated for fuel rods and burnable poisons containing no burnable poisons. A method for calculating core performance of a nuclear reactor, which is characterized in that the fuel rods are separately provided. 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 a nuclear 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 ratio of the thermal neutron flux to the fast neutron flux calculated using the distribution of the ratio of the thermal neutron flux to the fast neutron flux in the small volume given by the aggregate inhomogeneity calculation when the control rod is inserted in the small volume A method for calculating core performance of a nuclear reactor, which is characterized in that it is corrected by using an average burnup value of 13. 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 single set of the nuclear constants resulting from the small volume burning due to interaction of neutrons with adjacent small volume For changes from body combustion calculation, while using the burnup average value of the ratio of thermal neutron flux to fast neutron flux in the small volume, the control rod was inserted into the small volume and burned For the change from the single-assembly combustion calculation of the nuclear constant due to, the correction using the burnup integrated value during the period when the control rod was inserted and the burnup integrated value after the control rod was pulled out was performed. Add A method of calculating core performance of a nuclear reactor characterized by the above. 14. A correction coefficient according to a distance from the control rod is used as a correction coefficient according to a distance from the control rod, the variation amount of a nuclear constant in the small volume resulting from the insertion of the control rod into the small volume from a single-assembly combustion calculation. The core performance calculation method for a nuclear reactor according to claim 13, which is obtained by using the method. 15. The fuel in an assembly 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 average nuclear constant for the small volume given by a single assembly combustion calculation. In a reactor core performance calculation method for calculating a rod power distribution, the small volume corrects the fuel rod power in the small volume due to combustion due to the interaction of neutrons with the adjacent small volume, and the fuel A method for calculating core performance of a reactor characterized in that the local power of the rod is corrected by using the burn-up of local peaking of each fuel rod and the sensitivity coefficient to the burn-up mean value of the ratio of thermal neutron flux to fast neutron flux. . 16. Several fuel rods that are thermally severe in a small region of the core are selected as candidates, and for these candidates, changes in nuclear constants from single-assembly burnup calculations are considered. The method for calculating core performance of a nuclear reactor according to any one of claims 4 to 15, wherein the fuel rod output is calculated. 17. 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 single set of the nuclear constants resulting from the small volume burning due to interaction of neutrons with adjacent small volume A method for calculating core performance of a nuclear reactor, which comprises calculating a critical power ratio, which is a monitoring index for burnout of fuel rods, using fuel rod powers that take into account changes from body combustion calculations. 18. 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. Distribution and fuel rod power distribution in the assembly, in the core performance calculation method of a reactor, the change from the single assembly combustion calculation of the nuclear constant due to the combustion of the control rod inserted into the small volume A method for calculating core performance of a nuclear reactor, which comprises calculating a critical power ratio, which is a monitoring index for burnout of fuel rods, using the fuel rod power in consideration of the above.