JP2003161796A - Method and device for evaluating thermal characteristics of fuel rod - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a method and device for evaluating the thermal characteristics of fuel rods capable of evaluating, in detail, the thermal soundness of each fuel rod in a fuel assembly disposed in the reactor core of a boiling water reactor. <P>SOLUTION: A fuel assembly node output distribution in which a plurality of regions acquired by axially dividing a fuel assembly disposed in the reactor core of a boiling water type nuclear reactor are taken as unit nodes for calculation, is multiplied by a peaking factor for every fuel rod of the fuel assembly given a combustion degree and void ratio for every node as functions, so that the thermal characteristics of a unit node is calculated for all or specific fuel rod. The result is compared with a limit value set from the relationship with a unit node combustion degree of each fuel rod, to evaluate thermal soundness of every fuel rod. <P>COPYRIGHT: (C)2003,JPO

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a fuel rod thermal characteristic evaluation method and apparatus for evaluating the thermal integrity of each fuel rod of a fuel assembly arranged in the core of a boiling water reactor. A fuel rod thermal characteristic evaluation method that enables detailed calculation of the thermal characteristics of each fuel rod and detailed evaluation of the thermal soundness of each fuel rod from the relationship with the burnup of each fuel rod. And equipment.

[0002]

2. Description of the Related Art The core of a boiling water reactor is constructed as shown in FIG. In the core, a plurality of fuel assemblies 1
Is inserted, and the fuel assembly 1 of each is
The structure is such that 100 fuel rods are arranged in a matrix and covered with a channel box of about 15 cm square, and the effective length is about 3.7 m. A plurality of control rods 2 and a plurality of in-core neutron detectors 3 for detecting neutron flux are arranged in the core.

In order to ensure the integrity of the fuel during operation of the boiling water reactor, the heat output generated per unit length of the fuel rods constituting the fuel assembly 1 (hereinafter referred to as "linear power density"). ) Must be kept below the limit. In the conventional technique, in monitoring or predicting the core, it is general to calculate and evaluate the linear power density by the following method.

That is, first, each fuel assembly 1 forming the core is divided into a plurality of unit nodes in the axial direction, and it is approximated that the composition of the fuel is homogeneous in this unit node, so that the minority group or the modified 1 Solve the neutron diffusion equation of the group,
Calculate unit node average neutron flux and heat output (density). Next, the linear power density (maximum power output) of the fuel rod is calculated by multiplying the unit node average thermal power (density) by the peaking coefficient (“maximum peaking coefficient”) of the fuel rod with the highest thermal output in the fuel assembly. Calculate the linear power density "). The maximum peaking coefficient used here is generally calculated by a fuel assembly core calculation code and given as a function of the burnup and void fraction of the fuel assembly. It is evaluated that the thermal soundness of the fuel rods of the fuel assembly can be secured if the maximum linear power density obtained as described above is controlled to be equal to or less than the defined limit value.

The limit value used in the prior art is defined as a constant value or as a function of the burnup of the unit node of the fuel assembly, so that it can be easily evaluated.

[0006]

However, in the state where the fuel has been used up to a high burnup range like the recent fuel, the thermal soundness of the fuel rod is strong against the burnup of the fuel rod. I've learned to be affected. Therefore, as in the prior art, in the method of evaluating only the linear power density of the maximum output fuel rod and comparing it with the limit value that is not related to the burnup of the fuel rod, the soundness of the fuel rod can be accurately measured. There are drawbacks that cannot be evaluated. In addition, in the evaluation method of the prior art, it is general that the limit value that does not use the burnup of the fuel rod is comprehensively set and evaluated, and it is necessary to set and operate the limit value with an excessive margin. As a result, the economy of fuel is impaired.

As described above, the conventional evaluation method has a drawback that the soundness of the fuel rod cannot be accurately evaluated in the high burnup range. Further, in order to maintain the soundness of the fuel rod, it is necessary to set and operate a limit value with an excessive margin, which impairs fuel economy.

In view of the above, the present invention is a fuel rod thermal characteristic evaluation method capable of evaluating in detail the thermal integrity of each fuel rod of a fuel assembly arranged in the core of a boiling water reactor. And to obtain the device.

[0009]

According to a first aspect of the present invention, a fuel assembly calculated by using a region in which a fuel assembly arranged in a core of a boiling water reactor is divided into a plurality of units as a unit node is calculated. By multiplying the node output distribution by the peaking coefficient for each fuel rod of the fuel assembly given as a function of burnup and void fraction for each node, the thermal characteristics of the unit node can be determined for all or specific fuel rods. It is characterized in that the thermal soundness of each fuel rod is evaluated by calculating and comparing with a limit value determined from the relationship with the unit node burnup of each fuel rod.

According to a second aspect of the invention, in the first aspect of the invention, the signal of the fuel assembly node output distribution is utilized by utilizing the signal of the in-reactor neutron detector arranged in the core of the boiling water reactor. The feature is that the calculated value is corrected.

According to a third aspect of the invention, in the invention according to the first or second aspect, when the peaking coefficient for each fuel rod of the fuel assembly arranged in the core of the boiling water reactor is calculated, It is characterized in that the correction is performed by at least one of the change amount of the neutral flux due to the influence of one-sided combustion with the control rod inserted and the change amount of the neutron flux due to the influence of the adjacent fuel assembly.

The invention according to claim 4 relates to claims 1 to 3.
In the invention described in any one of, when calculating the unit node burnup for each fuel rod of the fuel assembly placed in the core of the boiling water reactor, the amount of change in neutron flux due to the influence of control rods and It is characterized in that it is corrected by integrating at least one of the changes in the neutron flux due to the influence of adjacent fuel assemblies.

According to a fifth aspect of the invention, in the invention according to any one of the first to fourth aspects, the fuel assembly node output distribution is stored in time series for a plurality of states of the boiling water reactor. Every other time, the peaking coefficient of each fuel rod of the fuel assembly is multiplied to evaluate the time change of the thermal characteristics of the fuel rods for all or a specific fuel rod.

The invention according to claim 6 relates to claims 1 to 5.
In the invention described in any one of the above, for a plurality of states of the boiling water reactor, by storing the fuel assembly node output distribution in time series, by multiplying the peaking coefficient for each fuel rod of the fuel assembly , All or specific fuel rods, the temporal change of the thermal properties of the fuel rods is evaluated, and the thermomechanical properties of the fuel rods are evaluated using the results.

The invention according to claim 7 relates to claims 1 to 6.
In the invention described in any one of 1, the program for calculating the fuel assembly node output distribution and the program for calculating the thermal characteristics of each fuel rod in the fuel assembly are separated and processed by the same or different computers. And

The invention according to claim 8 relates to claims 1 to 7.
In the invention described in any one of 1, the computer for calculating the fuel assembly node output distribution and the computer for calculating the thermal characteristics for each fuel rod in the fuel assembly are separated, and the thermal characteristics for each fuel rod in the fuel assembly are separated. It is characterized in that a plurality of computers for calculating characteristics are processed.

The invention according to claim 9 relates to claims 1 to 8.
In the invention described in any one of the above, among the fuel assembly node output distribution periodic calculation results of the core performance monitoring device installed in the nuclear power plant, all or only specific results are stored in the same computer or different computers. In addition, the thermal characteristics of each fuel rod in the fuel assembly are calculated.

According to a tenth aspect of the present invention, in the invention according to any one of the first to ninth aspects, the fuel assembly node output distribution periodic calculation result of the core performance monitoring device installed in the nuclear power plant is always calculated. Alternatively, it is characterized in that it is transmitted to different computers using a network at a specific cycle to calculate the thermal characteristics of each fuel rod in the fuel assembly.

The invention according to claim 11 is the invention according to any one of claims 1 to 10, in which all or all of the fuel assembly node output distribution periodic calculation results of the core performance monitoring device installed in the nuclear power plant are A part of it is saved, and the calculation result for a specific fuel assembly requested by using the network is transmitted to different computers, and the thermal characteristics of each fuel rod in the fuel assembly are calculated. And

According to a twelfth aspect of the present invention, there is provided a calculation condition setting section for setting a fuel rod for evaluating the operating state and thermal characteristics of the reactor to be evaluated and storing it in a data storage section, and the above calculation condition setting. For the operating conditions set in the section, the unit node average neutron flux, heat output, and void fraction distribution are calculated, and the fuel assembly node output distribution calculation unit is stored in the above data storage unit, and the unit node average heat output. And the void fraction distribution are used to calculate the unit node burnup and the history void fraction, and the fuel assembly node burnup distribution calculation unit that stores the data in the data storage unit and the unit node burnup and void fraction distribution are used to calculate the fuel The output power peaking coefficient for each rod is calculated and the linear output density for each fuel rod is calculated using the fuel rod output peaking coefficient calculation unit that stores it in the data storage unit, the unit node average heat output and the fuel rod output peaking coefficient. Fuel rod power density calculation unit that calculates the fuel rod burn-up density stored in the data storage unit, and calculates the burnup of each fuel rod using the unit node burnup and the history void ratio, and stores it in the data storage unit Section, the thermal limit value comparative evaluation section for performing comparative evaluation with the limit value for each fuel rod by using the linear power density for each fuel rod and the burnup for each fuel rod, and the thermal limit value for comparative evaluation. It is characterized in that it has a calculation result output unit for collecting the comparison result with the limit value for each fuel rod which is the output of the unit and outputting the information and detailed information of the severest fuel rod in the reactor.

According to a thirteenth aspect of the present invention, in the twelfth aspect of the present invention, an output distribution calculator for correcting the fuel assembly node output distribution is provided.

The invention according to claim 14 is characterized in that, in the invention according to claim 12 or 13, there is provided an output peaking coefficient correction calculation section for correcting the fuel rod output peaking coefficient.

According to a fifteenth aspect of the present invention, in the invention according to any one of the twelfth to fourteenth aspects, a fuel rod burnup correction calculation unit for correcting the burnup of the fuel rod is provided.

[0024]

BEST MODE FOR CARRYING OUT THE INVENTION Embodiments of the present invention will be described below. FIG. 1 is a block diagram showing the first embodiment of the present invention.

In FIG. 1, reference numeral 10 is a calculation condition setting unit, which sets a fuel rod for evaluating the operating state and thermal characteristics of the reactor to be evaluated and stores it in the data storage unit 11. In the data storage unit 11, the unit node average neutron flux calculated by the fuel assembly node output distribution calculation unit 12, the heat output (density) and void fraction distribution, and the fuel assembly node burnup distribution calculation unit 13 calculate Burned unit node burn-up, history void fraction, fuel rod output peaking coefficient calculator 14
The output peaking coefficient for each fuel rod calculated in step 1, the linear power density for each fuel rod calculated in the fuel rod linear power density calculation part 15, and the burnup for each fuel rod calculated in the fuel rod burnup calculation part 16. Are input and stored respectively.

That is, the fuel assembly node output distribution calculation unit 12 solves the neutron diffusion equation with respect to the operating state set by the calculation condition setting unit 10, and the unit node average neutron flux, heat output (density) and The void fraction distribution is calculated, and the unit node average neutron flux, heat output (density), and void fraction distribution are stored in the data storage unit 11. Further, the fuel assembly node burnup distribution calculation unit 13 calculates the unit node burnup and the history void ratio using the unit node average heat output and void ratio distribution calculated by the fuel assembly node output distribution calculation unit 12. Then, the unit node burnup and the history void fraction are stored in the data storage unit 11. Further, the fuel rod output peaking coefficient calculation unit 14 uses the unit node burnup calculated by the fuel assembly node burnup distribution calculation unit 13 and the void fraction distribution calculated by the fuel assembly node output distribution calculation unit 12. The output peaking coefficient for each fuel rod is calculated and stored in the data storage unit 11. The fuel rod power density calculation unit 15 uses the unit node average heat output calculated by the fuel assembly node power distribution calculation unit 12 and the fuel rod output peaking coefficient calculated by the fuel rod output peaking coefficient calculation unit 14 The linear power density for each fuel rod is calculated and stored in the data storage unit 11. Further, the fuel rod burnup calculation unit 16 calculates the burnup for each fuel rod using the unit node burnup and the history void ratio calculated by the fuel assembly node burnup distribution calculation unit 13, and the data storage unit 11 is stored.

Then, using the linear power density for each fuel rod and the burnup for each fuel rod, the thermal limit value comparison and evaluation section 17 performs a comparative evaluation with the limit value for each fuel rod, and the thermal limit is carried out. The result of comparison with the limit value for each fuel rod, which is the output of the value comparison / evaluation unit 17, is totaled by the calculation result output unit 18, and the strictest fuel rod information and detailed information in the reactor are output.

The core of the boiling water reactor to be evaluated is constructed as shown in FIG. 8 as in the prior art. A plurality of fuel assemblies 1 are inserted in the core, and each fuel assembly 1 has 50 to 100 fuel rods arranged in a matrix and is covered with a channel box of about 15 cm square. It has a structure and its effective length is about 3.7 m. A plurality of control rods 2 and a plurality of in-core neutron detectors 3 for detecting neutron flux are arranged in the core.

In this embodiment, a region where the fuel assembly 1 arranged in the core of a boiling water reactor is divided into a plurality of parts in the axial direction is divided into unit nodes. In a class boiling water reactor, there are 764 fuel assemblies 1, and if the number of axial divisions is 24, then 18
It becomes 336 unit nodes. Moreover, since the fuel assembly 1 of the boiling water reactor is composed of 50 to 100 fuel rods, unit nodes of about 900,000 to about 1.8 million fuel rods are the calculation targets. Therefore, in the calculation for all the fuel rods, the calculation time and the storage capacity of the data storage unit 11 may become too large. Therefore, the calculation condition setting unit 10 can be set to limit the fuel rods to be calculated, if necessary. For example, fuel assembly 1
There is a method of calculating only the 1/4 region by dividing the inside of the reactor into four and dividing it into eight by utilizing the fact that the are regularly and symmetrically arranged in the reactor. Further, the linear power density of the maximum power fuel rod is calculated by the same method as the conventional technique, and the detailed fuel rod linear power density calculation is performed for the fuel rods included in the severe upper nodes (for example, about 100 nodes) in the core. There is also a way to set to run. Furthermore, there is also a method of setting to perform detailed fuel rod power density calculation for a specially designated fuel assembly.

The fuel assembly node output distribution calculator 12
By approximating that the composition of the fuel is homogeneous within the unit node, a minority group (the neutron group is divided into a fast neutron group, an epithermal neutron group, and a small number of energy groups of thermal neutron groups), or a modified one group Solving the neutron diffusion equation (the neutron group is modified to consist of only one group of fast neutrons) and coupling the nuclear thermal-hydraulic characteristic calculation, unit node average neutron flux, thermal power (density) and void fraction Calculate the distribution.

Fuel assembly node burnup distribution calculator 13
Is calculated by multiplying the unit node average heat output by the time increment from the starting state to obtain the unit node burnup increment.
By adding to the unit node burnup of the calculation starting point of 1,
Calculate the unit node burnup at the time of interest. Also,
By multiplying the void fraction distribution by the time increment from the starting point state to obtain the historical void fraction distribution increment and adding it to the historical void fraction of the calculation starting point of the data storage unit 11, the historical void fraction at the target time point is calculated.

The fuel rod output peaking coefficient calculation unit 14
The unit node burnup and void fraction distribution are used to calculate the power peaking coefficient for each fuel rod. There are various methods for calculating the output peaking coefficient for each fuel rod, but in the present embodiment, a calculation example of an infinite lattice system will be shown as a comparatively simple method example. The fuel rod output peaking coefficient in a unit node is unique to the unit node, which is generally distinguished by the geometric shape of the fuel assembly in the unit node, the enrichment of each fuel rod, and the distribution of combustible poisons such as gadolinia. Organized by the index (“fuel type”). The output peaking coefficient for each fuel rod is a non-homogeneous neutron in an infinite lattice system with void fractions v0, v1, v2 ... for each of burnups e0, e1, e2 ..., by the fuel assembly design calculation performed for each fuel type. Since it is obtained as a solution of the diffusion equation, its value is stored in the data storage unit 11 in the form of a numerical table or a numerical table of coefficients of the fitting equation. As a fitting formula,
There is a method of storing each burnup ei (i = 1, 2, ...) By a function consisting of a polynomial of void fraction, and a method of storing by a function consisting of a polynomial of burnup e and a void fraction v. It should be noted that the numerical table and the numerical table of the coefficients of the fitting equation can be obtained in two cases, with and without the control rod 2 included in the unit node of interest.

The fuel rod linear power density calculation unit 15 uses the unit node average heat output and the fuel rod power peaking coefficient to calculate the linear power density for each fuel rod node by the following equation. LHGR (I, J, K) = α * P * LPF (I, J, K) …………… (1) where LHGR (I, J, K) is the unit of fuel rod (I, J). The linear output density P of the node K: average heat output LPF (I, J, K) of the fuel assembly unit node: output peaking coefficient α of the unit node K of the fuel rod (I, J): proportional coefficient.

The fuel rod burnup calculation unit 16 calculates the burnup for each fuel rod node by the following equation using the unit node burnup and the hysteresis void ratio. LEXP (I, J, K) ＝ E ＊ LPE (I, J, K) …………… (2) where LEXP (I, J, K) is the unit node K of the fuel rod (I, J). Burnup E of fuel assembly unit node LPE (I, J, K): burnup peaking coefficient of unit node K of fuel rod (I, J).

The burnup peaking coefficient: LPE (I, J, K) of the unit node K of the fuel rod (I, J) used in the above calculation is calculated by the fuel assembly design calculation performed for each fuel type. Every time
For each of e0, e1, e2 ... History void ratios vh0, vh1,
Since it is obtained as a solution of the non-homogeneous neutron diffusion equation in the infinite lattice system of vh2 ..., Its value is stored in the data storage unit 11 as a numerical table or a numerical table of coefficients of the fitting equation. As a fitting formula, burnup ei (i =
There is a method of storing with a function consisting of a polynomial of the hysteresis void ratio for each 1, 2, ...

The thermal limit value comparison / evaluation unit 17 determines the linear power density for each fuel rod node obtained by the equation (1): LHGR (I, J, K).
And the burnup for each fuel rod node obtained by equation (2): LEX
P (I, J, K) is used to perform a comparative evaluation with the limit value for each fuel rod node (“fuel rod thermal limit value”). The fuel rod thermal limit value is determined by various safety analyzes as in the example (reference numeral 19) shown in FIG. 2, and generally tends to decrease as the burnup of each fuel rod node increases. The thermal limit value comparison / evaluation unit 17 stores the fuel rod thermal limit value 19 as a numerical table or a function in relation to the fuel rod burnup, and the burnup for each fuel rod node: LEXP (I, J, Fuel rod thermal limit value corresponding to (K): LIMIT
Find (I, J, K). Obtained fuel rod thermal limit value: LIMIT
Using (I, J, K), the linear power density for each fuel rod node: LHGR (I, J, K) is evaluated by a ratio according to the following formula. FLPD (I, J, K) ＝ LHGR (I, J, K) / LIMIT (I, J, K) …………… (3) where FLPD (I, J, K): Fuel rod (I , J) to the limit value of the unit node K LHGR (I, J, K): linear power density LIMIT (I, J, K) of the unit node K of the fuel rod (I, J): fuel rod (I, J It is the fuel rod thermal limit value for the burnup of the unit node K of J).

The calculation result output unit 18 is a ratio of the output of the thermal limit value comparison / evaluation unit 17 to the limit value for each fuel rod:
FLPD (I, J, K) is totaled and the information of the severest fuel rod in the fuel assembly and the severest fuel rod in the reactor is output.

As described above, in the present invention, the unit node line power density for each fuel rod is calculated and compared with the fuel rod thermal limit value determined from the relationship with the unit node burnup for each fuel rod. By doing so, the thermal integrity of individual fuel rods can be accurately evaluated.

FIG. 3 is a block diagram showing a second embodiment of the present invention, in which an output distribution correction calculator 20 is added to the first embodiment. Generally, the fuel assembly node output distribution calculation unit 12 solves the neutron diffusion equation and calculates the average neutron flux and heat output (density) of the unit node, but includes a calculation error due to modeling. In the reactor core, it is possible to measure the neutron flux distribution in the fuel assembly gap from the local power monitor by multiple fixed neutron detectors and the reading of the movable neutron detector that moves inside the tube in the axial direction. An in-reactor neutron detector 3 is installed at. The power distribution correction calculation unit 20 compares the neutron flux distribution calculation value of the fuel assembly gap calculated by the fuel assembly node output distribution calculation unit 12 with the measurement value of the in-reactor neutron detector 3, thereby Calculate the power distribution correction coefficient that makes the neutron flux and heat output distribution of the device match the measured values. Then, using the output distribution correction coefficient, the solution of the neutron diffusion equation is corrected in the fuel assembly node output distribution calculation unit 12, and the calculation error due to modeling is corrected. In addition,
The above-mentioned power distribution correction coefficient corrects a calculation error caused by modeling, and has an advantage that it can be applied also when evaluating a future operating state of a nuclear reactor. Therefore, also in this embodiment, the same operational effect as that of the first embodiment is achieved while improving the calculation accuracy.

FIG. 4 is a block diagram showing a third embodiment of the present invention. An output peaking coefficient correction calculation section 21 is added to the first embodiment. In the first embodiment, the fuel rod output peaking coefficient calculation unit 1
In No. 4, when calculating the output peaking coefficient for each fuel rod using the unit node burnup and the void fraction distribution, the numerical table as the calculation result of the infinite grid system or the numerical table of the coefficient of the fitting formula was used. However, the neutron distribution in the fuel assembly placed in the reactor is affected by one-sided burning with the control rod 2 inserted and the adjacent fuel assembly, so the neutron distribution of the infinite lattice system is There will be some difference between and. Therefore, in order to obtain the output peaking coefficient of each fuel rod in the fuel assembly arranged in the nuclear reactor more accurately, the output peaking coefficient of the infinite lattice system should be adjusted to one It is necessary to consider the effects and the effects of adjacent fuel assemblies. The third embodiment is
The output peaking coefficient correction calculator 21 is added to perform the above correction. As an example of the calculation method of the output peaking coefficient correction calculation unit 21, the influence of one-sided combustion in the state where the control rods are inserted shows the change amount of the output peaking coefficient for each fuel rod, which is proportional to the control rod insertion period. There is a method of making corrections using a table of coefficients in the fitting formula. Further, as a method of considering the influence of adjacent fuel assemblies, the neutron flux of the fuel assembly of interest in the infinite lattice system and the neutron flux of the fuel assembly adjacent to the fuel assembly of interest Obtain the boundary value of the neutron diffusion equation, solve the neutron diffusion equation in the fuel assembly of interest using this boundary value, calculate the change amount from the neutron flux of the infinite lattice system, and calculate the correction amount of the output peaking coefficient There is a way.

When considering the influence of the adjacent fuel assemblies, due to the influence of the bending of the channel box,
In some cases, the effect of changing the water gap between the fuel assemblies is considered. The fuel rod output peaking coefficient calculation unit 14 calculates the output peaking coefficient for each fuel rod of the infinite lattice system using the unit node burnup and the void fraction distribution, and the output peaking correction calculation unit 21 calculates the output peaking coefficient. Multiply the coefficient correction amount to calculate the output peaking coefficient for each fuel rod. Therefore, the output peaking coefficient for each fuel rod in the fuel assembly can be more accurately determined in consideration of the influence of one-sided combustion with the control rod inserted and the influence of the adjacent fuel assembly. Therefore, also in this embodiment, the same operational effects as those of the first and second embodiments are achieved while improving the calculation accuracy.

FIG. 5 is a block diagram showing a fourth embodiment of the present invention, in which a fuel rod burnup correction calculation unit 22 is added to the first embodiment. In the first embodiment, the fuel rod burnup calculation unit 16 calculates the burnup for each fuel rod by using the unit node burnup and the history void fraction distribution, and the calculated number of the infinite lattice system. A table or a table of coefficients of the fitting formula was used. But,
Burnup peaking in a fuel assembly placed in a nuclear reactor is affected by the period during which the control rod 2 was inserted in the past, or by adjacent fuel assemblies. There will be some difference between and. Fourth
In the embodiment, the fuel rod burnup correction calculation unit 22 is added to perform the above correction. As a calculation method of the fuel rod burnup correction calculation unit 22, during the period when the control rod 2 is inserted, output peaking in the state where the control rod 2 is inserted is used, and the burnup peaking is corrected by integrating and accumulating. There is a way to calculate the quantity. In addition, by the same method as the method of calculating the correction amount of the output peaking coefficient described in the third embodiment, the influence of the adjacent fuel assemblies is calculated, the output peaking is calculated, and integrated and accumulated. The correction amount of burnup peaking can be calculated by. When considering the effects of adjacent fuel assemblies, the correction amount for burnup peaking is calculated by integrating the effect by the bending of the channel box that changes the water gap between the fuel assemblies. In some cases. The fuel rod burnup calculation unit 16 calculates burnup peaking for each fuel rod in the infinite lattice system using the unit node burnup and the history void fraction distribution, and the burnup calculated by the fuel rod burnup correction calculation unit 22. Multiply the degree peaking correction amount to calculate the burnup for each fuel rod. Therefore, it is possible to more accurately determine the burnup of the fuel rods in the fuel assembly in consideration of the influence of the control rod insertion period and the influence of the adjacent fuel assemblies. Therefore, also in this embodiment, the same operational effects as those of the first to third embodiments are achieved while improving the calculation accuracy.

FIG. 6 is a block diagram showing the fifth embodiment of the present invention. In the first embodiment, a plurality of data storage units 11 are provided. In the first embodiment, the operating state of the reactor to be evaluated is one, but in the fifth embodiment, a plurality of operating states of the reactor are the targets of evaluation. In the fifth embodiment, the fuel assembly node output distribution calculation unit 12 calculates the unit node average neutron flux for one reactor operating state among a plurality of reactor operating states to be evaluated. , The heat output (density) and the void ratio distribution are calculated, and the fuel assembly node burnup distribution calculation unit 13 calculates the unit node burnup and the history void ratio, and stores them in the data storage unit 11. Then, for the next reactor operating state, similarly, the fuel assembly node output distribution calculation unit 12 calculates the unit node average neutron flux, heat output (density) and void fraction distribution, and the fuel assembly node The burnup distribution calculation unit 13 calculates the unit node burnup and the history void ratio, and stores them in the data storage unit 11. The fuel assembly node output distributions of the plurality of operating states of the reactor calculated and stored as described above are arranged in time series, and the fuel rod output peaking coefficient calculation unit 14 calculates the same as in the first embodiment. , The unit node burnup and the void fraction distribution are used to calculate the output peaking coefficient for each fuel rod, and the fuel rod power density calculation unit 15 uses the unit node average heat output and the fuel rod output peaking coefficient to calculate the fuel. Calculate the linear power density for each bar. Further, the fuel rod burnup calculation unit 16 calculates the burnup for each fuel rod using the unit node burnup and the history void fraction. The thermal limit value comparison / evaluation unit 17 uses the linear output density of each fuel rod and the burnup of each fuel rod to perform a comparative evaluation with the limit value of each fuel rod. The calculation result output unit 18 outputs the time change of the thermal characteristics of the fuel rod calculated as described above. The above calculation can be applied to all fuel rods, but it is also possible to limit the evaluation to specific fuel rods. Therefore, also in this embodiment, it is possible to efficiently handle the time change of the thermal characteristics of the fuel rods, and the first to fourth
The same effects as those of the embodiment described above are obtained.

By the way, by using the time change of the thermal characteristics of the fuel rod obtained as described above, the behavior of the fuel rod is analyzed, and the heat of the fuel rod such as the gas pressure in the fuel rod and the cladding stress is measured. If the mechanical characteristics are evaluated, the soundness of the fuel rod can be evaluated more accurately.

FIG. 7 is a block diagram showing a sixth embodiment of the present invention. In the first embodiment, a data takeover unit 23 is added and calculation processing is separated. In the sixth embodiment, in the upstream side processing, the fuel assembly node output distribution calculation unit 12 determines the unit node average neutron flux, heat output (density) and voids for the operating state of the reactor to be evaluated. The rate distribution is calculated, and the fuel assembly node burnup distribution calculation unit 13 calculates the unit node burnup and the history void ratio, and stores them in the upstream data storage unit 11a. The data takeover unit 23 uses the upstream data storage unit 11a.
Data is stored in the downstream data storage unit 11b. In the downstream processing, the fuel rod output peaking coefficient calculation unit 14
Uses the unit node burnup and the void fraction distribution to calculate the output peaking coefficient for each fuel rod, and the fuel rod wire power density calculation unit 15 uses the unit node average heat output and the fuel rod output peaking coefficient to calculate Calculate the linear power density for each fuel rod.
Further, the fuel rod burnup calculation unit 16 calculates the burnup for each fuel rod using the unit node burnup and the history void fraction.
The thermal limit value comparison / evaluation unit 17 uses the linear output density of each fuel rod and the burnup of each fuel rod to perform a comparative evaluation with the limit value of each fuel rod. The calculation result output unit 18 collects the comparison result with the limit value for each fuel rod, and outputs information and detailed information of the severest fuel rod in the nuclear reactor. As described above, by performing the processing separately, it becomes possible to efficiently perform a large amount of data processing. It should be noted that it is possible to carry out the separated processing in a computer different from the method of processing in the same computer. Therefore, also in this embodiment, the processing can be efficiently performed, and the same effects as those of the first to fifth embodiments can be obtained.

If a plurality of computers for calculating the thermal characteristics of each fuel rod in the fuel assembly, which is the above-mentioned processing on the downstream side, are processed by a plurality of computers, a large amount of data processing can be efficiently carried out. . In particular, in the case of evaluating the thermomechanical properties of the fuel rod such as the gas pressure inside the fuel rod and the cladding stress by using the behavior analysis program of the fuel rod by using the temporal change of the thermal properties of the fuel rod, take time. Since this processing is a long-time processing that can be performed independently of each other, there is a great advantage in performing efficient processing using a plurality of computers.

By the way, the calculation of the fuel assembly node output distribution calculation unit 12 and the calculation of the fuel assembly node burnup distribution calculation unit 13, which are the processes on the upstream side, are performed by the core performance monitoring device installed in the nuclear power plant. This is the same calculation as the period calculation performed in. Therefore, of the periodic calculation results of the fuel assembly node output distribution performed by the core performance monitoring device installed at the nuclear power plant, all or only specific results are stored in the same computer or different computers, and the downstream side By calculating the thermal characteristics of each fuel rod in the fuel assembly,
Enables effective evaluation of safe operation of nuclear power plants.

Further, the periodic calculation result of the fuel assembly node output distribution performed by the core performance monitoring device installed in the nuclear power plant is transmitted to different computers using the network constantly or at a specific period, Effective evaluation is possible by calculating the thermal characteristics of each fuel rod in the assembly.

Furthermore, all or part of the fuel assembly node output distribution periodic calculation result of the core performance monitoring device installed in the nuclear power plant is saved and a specific fuel requested by using the network is stored. By transmitting the calculation results of the assembly to different computers and calculating the thermal characteristics of each fuel rod in the fuel assembly, a more effective evaluation is possible.

[0050]

As described above, according to the present invention, the thermal characteristics of each fuel rod of the fuel assembly arranged in the core of the boiling water reactor are calculated in detail, and the combustion of each fuel rod is performed. Since it is possible to compare with the limit value determined from the relationship with the degree, it is possible to accurately evaluate the health of the fuel rod, and it is possible to operate the nuclear power plant safely. Further, unlike the conventional case, it is not necessary to set and operate the limit value with an excessive margin, so that the economical efficiency of the fuel is not impaired and the economical efficiency of the nuclear power plant can be improved.

[Brief description of drawings]

FIG. 1 is a block diagram showing a first embodiment of a fuel rod thermal characteristic evaluation method of the present invention.

FIG. 2 is a diagram showing a thermal limit value of a fuel rod.

FIG. 3 is a block configuration diagram showing a second embodiment of a fuel rod thermal characteristic evaluation method of the present invention.

FIG. 4 is a block diagram showing a third embodiment of the fuel rod thermal characteristic evaluation method of the present invention.

FIG. 5 is a block configuration diagram showing a fourth embodiment of a fuel rod thermal characteristic evaluation method of the present invention.

FIG. 6 is a block configuration diagram showing a fifth embodiment of a fuel rod thermal characteristic evaluation method of the present invention.

FIG. 7 is a block diagram showing a sixth embodiment of the fuel rod thermal characteristic evaluation method of the present invention.

FIG. 8 is a sectional view of a core of a boiling water reactor.

[Explanation of symbols]

1 fuel assembly 2 control rod 3 In-core neutron detector 10 Calculation condition setting section 11 Data storage 12 Fuel assembly node output distribution calculator 13 Fuel assembly node burnup distribution calculator 14 Fuel rod output peaking coefficient calculator 15 Fuel rod power density calculator 16 Fuel rod burnup calculator 17 Thermal limit value comparison and evaluation section 18 Calculation result output section 19 Fuel rod thermal limit 20 Output distribution correction calculator 21 Output peaking coefficient correction calculator 22 Fuel rod burnup correction calculator 23 Data transfer part

   ─────────────────────────────────────────────────── ─── Continued front page    (72) Inventor Satoshi Hiratsuka             Kanagawa Prefecture Yokosuka City 23-13-1 Kawa             Ceremony company Global Nuclear Fue             Within Le Japan (72) Inventor Ken Toyonaga             Kanagawa Prefecture Yokosuka City 23-13-1 Kawa             Ceremony company Global Nuclear Fue             Within Le Japan (72) Inventor Tokunaga Kensuke             Kanagawa Prefecture Yokosuka City 23-13-1 Kawa             Ceremony company Global Nuclear Fue             Within Le Japan F term (reference) 2G075 AA03 CA08 DA18 FB05 FB07                       GA21

Claims (15)

[Claims] 1. A burnup and a void for each node in a fuel assembly node output distribution calculated with a unit node being a region obtained by dividing a fuel assembly arranged in a core of a boiling water reactor into a plurality of axially divided regions. Multiplying the peaking coefficient for each rod of the fuel assembly given as a function of rate, calculate the thermal properties of the unit node for all or certain rods, and burn up the unit node for each rod. The fuel rod thermal characteristic evaluation method is characterized by evaluating the thermal soundness of each fuel rod by comparing it with a limit value determined from the relationship. 2. A calculated value of a fuel assembly node output distribution is corrected by using a signal of an in-core neutron detector arranged in a core of a boiling water reactor. The fuel rod thermal characterization method described. 3. When calculating the peaking coefficient for each fuel rod of a fuel assembly arranged in the core of a boiling water reactor,
The correction is performed by at least one of the amount of change in the neutral flux due to the effect of one-sided combustion with the control rod inserted and the amount of change in the neutron flux due to the effect of the adjacent fuel assembly. 2. The fuel rod thermal characteristic evaluation method described in 2. 4. When calculating the unit node burnup for each fuel rod of a fuel assembly arranged in the core of a boiling water reactor, the amount of change in neutron flux due to the influence of control rods and adjacent fuel assemblies The fuel rod thermal characteristic evaluation method according to any one of claims 1 to 3, characterized in that the correction is performed by integrating at least one of the changes in the neutron flux due to the influence of the body. 5. A fuel assembly node output distribution for a plurality of states of a boiling water reactor is stored in time series, and the peaking coefficient of each fuel rod of the fuel assembly is multiplied to determine all or specific values. The fuel rod thermal characteristic evaluation method according to any one of claims 1 to 4, characterized in that the fuel rod thermal characteristic evaluation method according to claim 1 is evaluated with respect to time variation of thermal characteristics of the fuel rod. 6. A fuel assembly node output distribution for a plurality of states of a boiling water reactor is stored in time series, and the peaking coefficient for each fuel rod of the fuel assembly is multiplied to determine all or specific values. The fuel rod according to any one of claims 1 to 5, characterized in that the time change of the thermal characteristics of the fuel rod is evaluated, and the thermomechanical characteristics of the fuel rod are evaluated using the result. Bar thermal characterization method. 7. A program for calculating the fuel assembly node output distribution and a program for calculating the thermal characteristics of each fuel rod in the fuel assembly are separated and processed by the same or different computers. Item 7. The fuel rod thermal characteristic evaluation method according to any one of Items 1 to 6. 8. A computer for calculating the fuel assembly node output distribution and a computer for calculating the thermal characteristics of each fuel rod in the fuel assembly are separated, and the thermal characteristics of each fuel rod in the fuel assembly are calculated. The fuel rod thermal characteristic evaluation method according to claim 1, wherein a plurality of computers are used for processing. 9. A fuel assembly node output distribution periodic calculation result of a core performance monitoring device installed in a nuclear power plant,
The fuel according to any one of claims 1 to 8, wherein all or only specific results are stored in the same computer or different computers and the thermal characteristics of each fuel rod in the fuel assembly are calculated. Bar thermal characterization method. 10. A fuel assembly node output distribution periodic calculation result of a core performance monitoring device installed in a nuclear power plant is transmitted to different computers using a network at all times or at a specific cycle, and the fuel assembly is 10. The fuel rod thermal characteristic evaluation method according to claim 1, wherein the thermal characteristic of each fuel rod is calculated. 11. A fuel assembly node output distribution of a reactor core performance monitor installed in a nuclear power plant. All or part of a periodic calculation result is stored and a specific fuel requested by using a network is stored. The fuel rod thermal characteristic evaluation according to any one of claims 1 to 10, characterized in that the calculation results of the assembly are transmitted to different computers to calculate the thermal characteristics of each fuel rod in the fuel assembly. Method. 12. A calculation condition setting unit for setting a fuel rod for evaluating the operating state and thermal characteristics of a reactor to be evaluated and storing it in a data storage unit, and an operation set by the calculation condition setting unit. For the state, calculate the unit node average neutron flux, heat output and void fraction distribution, and use the fuel assembly node output distribution calculation unit to store in the above data storage unit, unit node average heat output and void fraction distribution Then, the unit node burnup and the history void fraction are calculated, and the fuel assembly node burnup distribution calculation unit that stores them in the data storage unit and the unit node burnup and void fraction distribution are used to output peaking coefficient for each fuel rod. Is calculated and stored in the data storage unit using the fuel rod output peaking coefficient calculation unit, the unit node average heat output and the fuel rod output peaking coefficient, and the linear power density for each fuel rod is calculated and stored in the data storage unit. The fuel rod power density calculation unit to be stored, the fuel rod burnup calculation unit that calculates the burnup of each fuel rod using the unit node burnup and the history void ratio, and stores it in the data storage unit; Using the linear power density and the burnup of each fuel rod, the output of the thermal limit value comparison and evaluation unit for performing comparative evaluation with the limit value for each fuel rod and the output of the thermal limit value comparison and evaluation unit A fuel rod thermal characteristic evaluation device having a calculation result output unit for collecting comparison results with respective limit values and outputting the strictest fuel rod information and detailed information in a nuclear reactor. 13. The fuel rod thermal characteristic evaluation device according to claim 12, further comprising an output distribution calculation unit that corrects a fuel assembly node output distribution. 14. The fuel rod thermal characteristic evaluation device according to claim 12, further comprising an output peaking coefficient correction calculation unit for correcting the fuel rod output peaking coefficient. 15. The fuel rod thermal characteristic evaluation device according to claim 12, further comprising a fuel rod burnup correction calculation unit for correcting the burnup of the fuel rod.