JP2004333433A - Method and device for evaluating thermal characteristic of fuel assembly - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To evaluate in detail thermal soundness of a fuel assembly arranged in a reactor core of a boiling water reactor. <P>SOLUTION: One or both of a linear power density of the fuel assembly and a critical power ratio thereof is (are) calculated using a peaking factor calculated using a burnup in the fuel assembly arranged in the reactor core of the boiling water reactor, a void ratio therein and a water gap width between channel boxes. The thermal soundness of the fuel assembly is evaluated based on the one or both of the linear power density of the fuel assembly and the critical power ratio thereof. <P>COPYRIGHT: (C)2005,JPO&NCIPI

Description

[0001]
TECHNICAL FIELD OF THE INVENTION
The present invention relates to a fuel assembly thermal characteristic evaluation method and apparatus for evaluating the thermal integrity of a fuel assembly disposed in a core of a boiling water reactor.
[0002]
[Prior art]
The core of the boiling water reactor is configured as shown in FIG. A plurality of fuel assemblies 1, a plurality of control rods 2, and a plurality of in-core neutron detectors 3 for detecting a neutron flux are arranged in the core. Each fuel assembly 1 is configured as shown in FIG. 8, in which 50 to 100 fuel rods 4 and several water rods 5 are arranged in a matrix and covered with a channel box 6 of about 15 cm square. The effective length is about 3.7 m.
[0003]
In order to ensure the integrity of the fuel assembly 1 during operation of the boiling water reactor, the generated heat output per unit length of the fuel rods 2 constituting the fuel assembly 1 (hereinafter referred to as “linear power density”) Call) must be kept below the limit value. In addition, even when an abnormal transient change occurs during operation, a limit output and a normal operation are set so that an output that causes fuel rod damage due to insufficient cooling (hereinafter referred to as “limit output”) does not occur. It is necessary that the ratio to the heat output of the fuel assembly 1 at the time (hereinafter, referred to as “limit output ratio”) be equal to or more than the limit value.
[0004]
In the prior art, in monitoring or predicting the core, it is general to calculate and evaluate the linear power density and the critical power ratio by the following method.
[0005]
That is, first, each fuel assembly 1 constituting the core is divided into a plurality of unit nodes in the axial direction, and within this unit node, the composition of the fuel is approximated to be homogeneous, so that the neutrons in the minority group or the modified group 1 are obtained. Solve the diffusion equation and calculate the unit node average neutron flux and heat output (density). Next, by multiplying the peak power coefficient (“local peaking coefficient”) of the fuel rod with the highest heat output in the fuel assembly by the heat output (density) of the unit node average, the linear power density of the fuel rod that is the maximum power is obtained. calculate. If the linear output density obtained as described above is controlled to be equal to or less than a predetermined limit value, it is evaluated that the thermal integrity of the fuel assembly 1 can be ensured.
[0006]
Further, the limit output is evaluated using an empirical formula obtained from an experiment in which an electric heating tube simulating the fuel rod 4 is configured in the shape of the actual fuel assembly 1. Since the empirical formulas are arranged using a coolant flow rate, a pressure, and a fuel assembly form factor data and a factor related to a peaking coefficient called an R factor, the coolant flow rate, pressure, and shape of each fuel assembly 1 are determined. The limit output is determined by the factor and the R factor. The critical output ratio is calculated as a ratio between the critical output of each fuel assembly 1 and the actual generated heat output. If the critical output ratio obtained as described above is controlled so as to be equal to or greater than a predetermined limit value, the thermal integrity of the fuel assembly 1 is evaluated as being assured.
[0007]
The above calculation requires a local peaking coefficient for linear power density calculation and an R factor for limiting power ratio calculation as a peaking coefficient for each fuel assembly 1. The local peaking coefficient and the R factor are calculated by the unit fuel assembly nucleus calculation code, assuming that the fuel assemblies 1 having the same shape are regularly arranged, and the burnup and the void of the fuel assembly 1 are calculated. It is common to give it as a function of rate. FIG. 9 is a schematic diagram showing a regular arrangement of the central lattice type fuel assemblies 1, in which the water gap widths (W) between the channel boxes of each fuel assembly are all equal. FIG. 10 is a schematic view showing a regular arrangement of the eccentric lattice type fuel assemblies 1, and the width (Ww) of the control rod insertion side of the water gap width between the channel boxes of each fuel assembly is controlled. Although the grid type is wider than the width (Wn) of the rod non-insertion side, the value of Ww is equal and the value of Wn is equal for all fuel assemblies.
[0008]
Since the peaking coefficient used in the conventional technique is calculated assuming that fuel assemblies having the same shape are regularly arranged as shown in FIG. 9 or FIG. 10, evaluation can be easily performed. ing.
[0009]
[Patent Document]
Japanese Patent Application No. 2001-360665
[0010]
[Problems to be solved by the invention]
However, in future fuel assemblies, use of channel boxes having different shapes has been considered. Normally, not all the fuel assemblies are replaced during one periodic inspection, but one-fifth to one-third replacement, so that a channel box having a different shape exists in the same core. As described above, the method using the peaking coefficient calculated on the assumption that the fuel assemblies having the same shape are regularly arranged has a disadvantage that the thermal integrity of the fuel assembly cannot be accurately evaluated. In addition, in the conventional evaluation method, even when using a channel box having the same shape, the manufacturing tolerance of the channel box and deformation during use cannot be taken into account. It must be defined and operated, which impairs fuel economy.
[0011]
As described above, the conventional evaluation method has a disadvantage that when using channel boxes having different shapes, the thermal integrity of the fuel assembly cannot be accurately evaluated. In addition, even when using channel boxes having the same shape, it is necessary to determine and operate a limit value with an excessive margin, which impairs fuel economy.
[0012]
In view of the above, the present invention provides a fuel assembly thermal characteristic evaluation method and apparatus capable of evaluating the thermal integrity of a fuel assembly disposed in a core of a boiling water reactor in detail. Aim.
[0013]
[Means for Solving the Problems]
According to the first aspect of the present invention, there is provided a fuel assembly using a burn-up, a void rate, and a peaking coefficient calculated using a water gap width between channel boxes in a fuel assembly disposed in a core of a boiling water reactor. One or both of the linear power density and the critical power ratio of the body are calculated, and the thermal integrity of the fuel assembly is evaluated based on one or both of the linear power density and the critical power ratio.
[0014]
The invention according to claim 2 is characterized in that, in the invention according to claim 1, the peaking coefficient is calculated as a function of the burnup, the void ratio, and the width of the water gap between the channel boxes.
[0015]
According to a third aspect of the present invention, in the invention according to the first aspect, the burnup of the fuel assemblies arranged in the core of the boiling water reactor is such that the channel boxes are regularly arranged in the same shape. The peaking coefficient of the fuel assembly is calculated by multiplying the peaking coefficient calculated as a function of the void fraction by the peaking correction amount for the irregular water gap width between the channel boxes, and calculating the linear power density of the fuel assembly. And one or both of the critical power ratio and the critical power ratio are evaluated.
[0016]
According to a fourth aspect of the present invention, in the invention according to any one of the first to third aspects, for a fuel assembly disposed in the core of a boiling water reactor, the water gap width between the channel boxes is set to each individual fuel. The calculation is performed using the design data of the aggregate or the measurement data before use.
[0017]
According to a fifth aspect of the present invention, in the invention according to any one of the first to fourth aspects, the water gap width between the channel boxes for each of the fuel assemblies disposed in the core of the boiling water reactor is set to an individual value. It is characterized in that the calculation is performed by adding a deformation amount due to elongation due to neutron irradiation to the design data of the fuel assembly or the measurement data before use.
[0018]
According to a sixth aspect of the present invention, in the invention according to any one of the first to fifth aspects, a neutron flux of a fuel assembly disposed in a core of a boiling water reactor is affected by an adjacent fuel assembly. The peaking coefficient is corrected based on at least one of the change amount and the change amount of the neutral flux due to the influence of one-sided burning.
[0019]
The invention according to claim 7 is the invention according to any one of claims 1 to 6, wherein a peaking coefficient of all or a specific fuel rod is calculated for a fuel assembly disposed in the core of the boiling water reactor. The maximum value is used as the peaking coefficient of the fuel assembly.
[0020]
The invention according to claim 8 is the invention according to any one of claims 1 to 7, wherein the peaking coefficient of all or a specific fuel rod is calculated for the fuel assembly disposed in the core of the boiling water reactor. Using the result obtained, an R factor for calculating the limit power ratio is calculated.
[0021]
According to a ninth aspect of the present invention, in the invention according to any one of the first to eighth aspects, a signal of a neutron detector in a reactor is used for a fuel assembly disposed in a core of a boiling water reactor. And correcting the calculated value of the node output distribution of the fuel assembly.
[0022]
According to a tenth aspect of the present invention, there is provided a calculation condition setting unit that sets an operation state of a boiling water reactor to be evaluated and stores the operation state in a data storage unit, and an operation state set in the calculation condition setting unit. A node output calculator for calculating a unit node average neutron flux, a heat output, and a void fraction distribution using a region obtained by dividing the fuel assembly into a plurality in the axial direction as a unit node, and storing the calculated data in the data storage unit; Using the node average heat output and the void fraction distribution, calculate the unit node burnup, calculate the node burnup calculation unit stored in the data storage unit, and calculate the water gap width between the channel boxes of the individual fuel assemblies, The peak gap coefficient of the fuel assembly is calculated using the water gap width calculation unit stored in the data storage unit, the unit node burnup, the void fraction distribution, and the water gap width, and stored in the data storage unit. A peaking coefficient calculator, a thermal characteristic calculator for calculating one or both of a linear power density and a critical power ratio of a fuel assembly using a unit node average heat output, a void fraction distribution and a peaking coefficient, and a calculation result And a calculation result output unit that outputs information and detailed information of the most severe fuel assemblies in the reactor.
[0023]
According to an eleventh aspect of the present invention, in the invention according to the tenth aspect, a unit node peaking coefficient calculating unit for calculating a peaking coefficient in a state where channel boxes are regularly arranged in the same shape as a function of burnup and void fraction. And a peaking coefficient calculator having a water gap width dependent peaking correction calculator for calculating a peaking correction amount due to an irregular water gap width between channel boxes.
[0024]
According to a twelfth aspect of the present invention, in the invention according to the tenth or eleventh aspect, there is provided a channel box deformation amount calculation unit for calculating a deformation amount caused by elongation due to neutron irradiation of the channel box.
[0025]
According to a thirteenth aspect of the present invention, in the invention according to any one of the tenth to twelfth aspects, at least one of a change amount of the neutron flux due to the influence of the adjacent fuel assembly and a change amount of the neutron flux due to the influence of the one-sided burning. A peaking coefficient correction calculator for correcting a peaking coefficient is provided.
[0026]
According to a fourteenth aspect of the present invention, in the invention according to any one of the tenth to thirteenth aspects, a node output correction calculation for correcting a calculated value of a node output distribution of a fuel assembly using a signal of a reactor neutron detector. It has a part.
[0027]
BEST MODE FOR CARRYING OUT THE INVENTION
Hereinafter, embodiments of the present invention will be described. FIG. 1 is a block diagram showing a first embodiment of the present invention.
[0028]
In FIG. 1, reference numeral 10 denotes a calculation condition setting unit. When the operation condition of the reactor to be evaluated is set by the calculation condition setting unit 10, the set operation condition is stored in the data storage unit 11. . The data storage unit 11 stores the neutron flux, heat output (density) and void fraction distribution of the unit node average calculated by the node output calculation unit 12, the unit node burnup calculated by the node burnup calculation unit 13, and the water. The water gap width calculated by the gap width calculator 14 and the peaking coefficient calculated by the peaking coefficient calculator 15 are input and stored.
[0029]
That is, the node output calculation unit 12 solves the neutron diffusion equation for the operation state set by the calculation condition setting unit 10, and calculates the neutron flux, heat output (density), and void fraction distribution of the unit node average, The neutron flux, heat output (density), and void fraction distribution of the unit node average are stored in the data storage unit 11. The node burnup distribution calculation unit 13 calculates the unit node burnup using the unit node average heat output calculated by the node output calculation unit 12, and stores the unit node burnup in the data storage unit 11. You. Further, the water gap width calculation unit 14 calculates the water gap width between the channel boxes of each fuel assembly, and stores the water gap width in the data storage unit 11. The output peaking coefficient calculator 15 calculates the unit node burnup calculated by the node burnup calculator 13, the void fraction distribution calculated by the node output calculator 12, and the water gap width calculated by the water gap width calculator 14. The peaking coefficient of the fuel assembly is calculated using the calculated value and stored in the data storage unit 11.
[0030]
The thermal characteristic calculation unit 16 uses the unit node average heat output calculated by the node output calculation unit 12 and the peaking coefficient calculated by the peaking coefficient calculation unit 15 to calculate the linear output density and the critical output ratio of the fuel assembly. The calculated result is summed up in the calculation result output unit 17, and information and detailed information of the strictest fuel assemblies in the reactor are output.
[0031]
The core of the boiling water reactor to be evaluated is configured as shown in FIG. 7, similarly to the prior art. A plurality of fuel assemblies 1, a plurality of control rods 2, and a plurality of in-core neutron detectors 3 for detecting a neutron flux are arranged in the core. As shown in FIG. 8, each fuel assembly 1 has a structure in which 50 to 100 fuel rods 4 and several water rods 5 are arranged in a matrix and covered by a channel box 6 of about 15 cm square. The effective length is about 3.7 m.
[0032]
In this embodiment, the fuel assembly 1 disposed in the core of the boiling water reactor is divided into a plurality of unit nodes in the axial direction. As a typical example, in a 1.1 million-class boiling water reactor, Assuming that there are 764 fuel assemblies 1 and the number of divisions in the axial direction is 24, there are 18336 unit nodes.
[0033]
The node output calculation unit 12 divides the neutron group into a small number of energy groups of a fast neutron group, an epithermal neutron group, and a thermal neutron group by approximating that the fuel composition is homogeneous within the unit node. Neutron diffusion equation and modified one group (a neutron group modified to consist of only a group of fast neutrons), coupled with nuclear thermo-hydraulic property calculations, and neutron flux and heat Calculate the output (density) and void fraction distribution.
[0034]
The node burn-up calculation unit 13 multiplies the unit node average heat output by the time increment from the starting state to obtain a unit node burn-up increment, and adds this to the unit node burn-up of the calculation start point of the data storage unit 11 to obtain the target Calculate the unit node burnup at the point in time.
[0035]
The water gap calculation unit 14 determines the position of the channel box 6 of the fuel assembly 1 to be calculated (the center fuel assembly in FIG. 2) for the irregular arrangement of the fuel assemblies 1 as shown in FIG. The water gap width (Wa, Wb, Wc, Wd) is calculated as an interval between the adjacent fuel assembly 1 and the channel box 6 for each surface. As the geometric shape of the channel box 6 used when calculating the water gap width, there are a method using design data relating to the shape and installation position of each channel box and a method using measurement data before use. Since the channel box 6 has a height of about 4 m and may have a bend in the middle, if the bend data at the axial position exists as measurement data before use, the channel box 6 at the unit node position may be used. Using the measured data as a bend, the water gap width can be calculated in detail for each axial position.
[0036]
The peaking coefficient calculation unit 15 calculates the peaking coefficient and the R factor by using the following formulas (1) and (2) using the unit node burnup, the void ratio, and the water gap width.
LPF = fl (EN, VF, Wa, Wb, Wc, Wd) (1)
RF = fr (EB, Wa, Wb, Wc, Wd) (2)
here,
LPF: Local peaking coefficient of unit node
RF: R factor of fuel assembly
EN: Unit node burnup
VF: void ratio of unit node
EB: Burnup of fuel assembly
Wa, Wb, Wc, Wd: Water gap width
fl: function for calculating local peaking coefficient
fr: Function for calculating R factor
[0037]
There are various methods for creating a function for calculating the local peaking coefficient and a function for calculating the R factor used in the above equation. In the present embodiment, the calculation is performed in advance using the unit fuel assembly calculation code. Here is an example.
[0038]
The local peaking coefficient in a unit node is an index unique to the unit node, which is generally distinguished by the geometry of the fuel assembly in the unit node of interest, the enrichment of each fuel rod, the distribution of burnable poisons such as gadolinia, etc. (Fuel type). .., (Wb1, Wb2, Wb3 ...), (Wc1, Wc2, Wc3 ...), (Wd1, Wd2, Wd3 ...) , And for each of the burn-ups E1, E2, E3,... And the void fractions V1, V2, V3,. The data is stored in the data storage unit 11 in the form of a numerical table or a numerical table of coefficients of a fitting equation. As the fitting equation, there is a method of storing a function composed of a polynomial of the void rate for each burnup Ei (i = 1, 2,...) And a method of storing a function of a polynomial relating to the burnup E and the void rate V. . It should be noted that the number table and the number table of coefficients of the fitting equation are obtained in two cases, with and without the control rod 2 in the unit node of interest, and are stored in each case.
[0039]
The R factor of the fuel assembly is arranged by an index (bundle type) unique to the fuel assembly, which is distinguished by the geometric shape of the fuel assembly, the enrichment of each fuel rod, the distribution of burnable poisons such as gadolinia, and the like. You. For each bundle type, the water gap width is set to a plurality of states (Wa1, Wa2, Wa3 ...), (Wb1, Wb2, Wb3 ...), (Wc1, Wc2, Wc3 ...), (Wd1, Wd2, Wd3 ...) Is calculated in advance for each of the burn-ups E1, E2, E3,... Using the unit fuel assembly calculation code, and the value is expressed in a numerical table or a numerical table of coefficients of a fitting equation. Is stored in the data storage unit 11. It should be noted that the number table and the number table of the coefficients of the fitting equation are stored in the form of a number table or a number table of the coefficients of the fitting equation according to the rate at which the control rod 2 is inserted into the fuel assembly of interest.
[0040]
In the function for calculating the local peaking coefficient and the function for calculating the R factor, there are various other methods for handling the water gap width. For example, instead of using individual water gap widths as they are, there is a method in which channel boxes are arranged based on deviations from design data when channel boxes are regularly arranged in the same shape. Further, there is a method of processing the absolute value of the deviation with the maximum value of the four surfaces, and a method of processing the absolute value of the deviation with the average of the four surfaces. Further, in consideration of the symmetry of the design of the fuel assembly, processing is performed with the maximum value of Wa and Wb and the maximum value of Wc and Wd, or with the average value of Wa and Wb and the average value of Wc and Wd. There is a way.
[0041]
The thermal characteristic calculation unit 16 calculates the linear output density for each unit node by using the following equation using the unit node average heat output and the local peaking coefficient.
LHGR = α * PN * LPF (3)
here,
LHGR: linear output density of unit node
PN: Average heat output of unit node
LPF: Local peaking coefficient of unit node
α: Proportional coefficient
[0042]
Further, the thermal characteristic calculation unit 16 calculates the critical output ratio of the fuel assembly by the following equation using the average heat output of the fuel assembly and the R factor.
CPR = CP / PB (4)
CP = fc (FB, RF, PR, β) (5)
here,
CPR: Fuel assembly critical power ratio
CP: Limit output of fuel assembly
PB: Average heat output of fuel assembly
fc: empirical formula for calculating the limit output
FB: Coolant flow rate of fuel assembly
RF: R factor of fuel assembly
PR: Reactor pressure
β: Shape factor of fuel assembly
[0043]
The calculation result output unit 17 sums up the linear power density and the critical power ratio, which are the outputs of the thermal characteristic calculation unit 16, and outputs the severest values in the core and information on individual fuel assemblies.
[0044]
As described above, in the present invention, the linear power density and the critical power ratio of the fuel assembly are calculated using the peaking coefficient calculated as a function of the burnup, the void fraction, and the water gap width between the channel boxes. Is compared with, for example, a limit value, the thermal integrity of the fuel assembly can be accurately evaluated.
[0045]
FIG. 3 is a block diagram showing a second embodiment of the present invention. The peaking coefficient calculator 15 according to the first embodiment includes a unit cell peaking calculator 15a and a water gap width dependent peaking correction calculator. 15b. The unit lattice peaking coefficient calculation unit 15a calculates the peaking coefficient as a function of the burnup and the void ratio in a state where the channel boxes are regularly arranged in the same shape (unit nodes). Further, the water gap width dependent peaking correction calculation unit 15b calculates a peaking correction amount due to the irregular water gap width between the channel boxes. That is, the peaking coefficient calculator 15 calculates the peaking coefficient by the following equation.
LPF = fld (EN, VF) * flg (VF, Wa, Wb, Wc, Wd, LIJ) (1a)
RF = frd (EB) * frg (Wa, Wb, Wc, Wd, RIJ) (2a)
here,
LPF: Local peaking coefficient of unit node
RF: R factor of fuel assembly
EN: Unit node burnup
VF: void ratio of unit node
EB: Burnup of fuel assembly
Wa, Wb, Wc, Wd: Water gap width
LIJ: Position of fuel rod generating local peaking coefficient
RIJ: Position of R-factor generating fuel rod
fld: function for calculating the local peaking coefficient of the unit cell
flg: function for correcting the local peaking coefficient depending on the water gap width
frd: function for calculating the R factor of the unit cell
frg: function for correcting the R factor depending on the water gap width
[0046]
The local peaking coefficient of the unit cell is generally determined by the unique index of the unit node, which is distinguished by the geometric shape of the fuel assembly in the unit node of interest, the enrichment of each fuel rod, the distribution of burnable poisons such as gadolinia, etc. "Fuel type"). For each of the fuel types, a local peaking coefficient is calculated in advance for each of the burnups E1, E2, E3,... For each of the void fractions V1, V2, V3,. The value is stored in the data storage unit 11 in the form of a number table or a number table of coefficients of a fitting equation. As the fitting equation, there is a method of storing a function composed of a polynomial of the void rate for each burnup Ei (i = 1, 2,...) And a method of storing a function of a polynomial relating to the burnup E and the void rate V. . It should be noted that the number table and the number table of coefficients of the fitting equation are obtained in two cases, with and without the control rod 2 in the unit node of interest, and are stored in each case.
[0047]
The function for correcting the local peaking coefficient depending on the water gap width includes the water gap width in a plurality of states (Wa1, Wa2, Wa3 ...), (Wb1, Wb2, Wb3 ...), (Wc1, Wc2, Wc3 ...), Assuming (Wd1, Wd2, Wd3...), A local peaking coefficient is calculated for a fuel assembly having a uniform enrichment using a unit fuel assembly calculation code, and the channel box has the same shape and regularity. The sensitivity function for each fuel rod position is determined in advance by comparing with the local peaking coefficient of the state (unit node) arranged in the. The above sensitivity function is calculated for a plurality of void ratios V1, V2, V3,... And stored in the data storage unit 11 in the form of a numerical table or a numerical table of coefficients of a fitting equation. It should be noted that the number table and the number table of coefficients of the fitting equation are obtained in two cases, with and without the control rod 2 in the unit node of interest, and are stored in each case.
[0048]
The R factor of the unit cell is arranged by an index (bundle type) unique to the fuel assembly, which is distinguished by the geometric shape of the fuel assembly, the enrichment of each fuel rod, the distribution of burnable poisons such as gadolinia, and the like. . For each of the bundle types, the R factor is calculated in advance for each of the burn-ups E1, E2, E3,... Using the unit fuel assembly calculation code, and the value is calculated as a numerical table or a numerical table of coefficients of a fitting equation. The format is stored in the data storage unit 11. It should be noted that the number table and the number table of the coefficients of the fitting equation are stored in the form of a number table or a number table of the coefficients of the fitting equation according to the rate at which the control rod 2 is inserted into the fuel assembly of interest.
[0049]
The function for correcting the R factor depending on the water gap width includes a plurality of states (Wa1, Wa2, Wa3 ...), (Wb1, Wb2, Wb3 ...), (Wc1, Wc2, Wc3 ...), (Wc1, Wc2, Wc3 ...), Assuming Wd1, Wd2, Wd3 ...), for a fuel assembly having a uniform enrichment, the R factor is calculated using the unit fuel assembly calculation code, and the channel boxes are regularly arranged in the same shape. The sensitivity function for each fuel rod position is determined in advance by comparing with the R factor in the set state (unit cell). The sensitivity function described above is stored in the data storage unit 11 as a numerical table according to the ratio of the control rods 2 inserted into the fuel assembly of interest or a numerical table of coefficients of a fitting equation.
[0050]
In the function for correcting the local peaking coefficient depending on the water gap width and the function for correcting the R factor depending on the water gap width, there are various other methods for handling the water gap width. For example, instead of using individual water gap widths as they are, there is a method in which channel boxes are arranged based on deviations from design data when channel boxes are regularly arranged in the same shape. Further, there is a method of processing the absolute value of the deviation with the maximum value of the four surfaces, and a method of processing the absolute value of the deviation with the average of the four surfaces. Further, in consideration of the compatibility of the design of the fuel assembly, a method of processing with the maximum value of Wa and Wb and the maximum value of Wc and Wd, and a method of processing with the average value of Wa and Wb and the average value of Wc and Wd There is a way.
[0051]
In addition, the peaking coefficient and the R factor are stored in the data storage unit 11 for all or a specific fuel rod in the fuel assembly of the unit cell, and correction is made for the irregular water gap width between the channel boxes. There is also a method of determining the local peaking coefficient and R factor of the fuel assembly by obtaining the maximum value in the fuel assembly based on the result.
[0052]
Further, the peaking coefficient is stored in the data storage unit 11 for all or specific fuel rods in the fuel assembly of the unit cell, and the result of the correction for the irregular water gap width between the channel boxes is shown. There is also a method of determining the R factor of the fuel assembly by calculating the R factor using the calculated values and calculating the maximum value in the fuel assembly.
[0053]
Thus, also in this embodiment, the same operation and effect as in the first embodiment can be obtained while reducing the required data amount.
[0054]
FIG. 4 is a block diagram showing a third embodiment of the present invention, in which a channel box deformation amount calculation unit 18 is added to the first embodiment. The channel box deformation amount calculation unit 18 calculates a bending amount of the channel box based on deformation caused by elongation due to neutron irradiation of the channel box. The water gap width calculation unit 14 adds the bending amount of the channel box calculated by the channel box deformation amount calculation unit 18 to the design data of the individual fuel assemblies or the measurement data before use, and calculates the water gap between the channel boxes. Calculate the width.
[0055]
The channel box deformation amount calculation unit 18 calculates the amount of elongation of the four surfaces of the channel box in the longitudinal direction using the neutron irradiation amounts of the four surfaces of the channel box stored in the data storage unit 11, and calculates the difference in elongation. Calculate the bending amount of the channel box from. For this calculation, the neutron flux on the four surfaces of the channel box is calculated when the node output calculation unit 12 calculates the node output, and stored in the data storage unit 11. Further, when the node burn-up calculation unit 13 calculates the node burn-up, the neutron doses on the four surfaces of the channel box are calculated and stored in the data storage unit 11.
[0056]
As another method, the channel box deformation amount calculation unit 18 uses the node burnup stored in the data storage unit 11 to calculate the neutron irradiation of the four surfaces of the channel box from the difference between the burnup increments of the adjacent fuel assemblies. The gradient of the amount is obtained, the difference in elongation of the channel box is calculated from the gradient, and the bending amount of the channel box is calculated.
[0057]
As described above, in the present embodiment, the peaking coefficient in the fuel assembly can be more accurately obtained using the water gap width in consideration of the deformation of the channel box. Thus, also in this embodiment, the same operational effects as those of the first and second embodiments can be obtained while improving the calculation accuracy.
[0058]
FIG. 5 is a block diagram showing a fourth embodiment of the present invention, in which a peaking coefficient correction calculator 19 is added to the first embodiment. In the first embodiment, when calculating the peaking coefficient, the peaking coefficient calculation unit 15 uses a numerical table, which is a calculation result of the unit cell system, or a numerical table of the coefficients of the fitting equation. However, the neutron distribution in the fuel assembly placed in the reactor is affected by the one-sided combustion with the control rod 2 inserted and the influence of the adjacent fuel assembly. And a slight difference occurs. Therefore, in order to more accurately determine the peaking coefficient in the fuel assembly placed in the reactor, the peaking coefficient of the unit cell system is affected by the one-sided burning with the control rod inserted and the fuel assembly adjacent to the fuel cell. The effects of the body need to be considered. In the fourth embodiment, a peaking coefficient correction calculator 19 is added to perform the above correction. As an example of the calculation method of the peaking coefficient correction calculator 19, the influence of one-sided burning in a state where the control rod is inserted is obtained by calculating the amount of change of the peaking coefficient in proportion to the control rod insertion period by the number table or the number of coefficients in the fitting equation. There is a correction method using a table. In addition, as a method of considering the influence of the adjacent fuel assembly, the neutron flux of the fuel assembly of interest in the unit cell system and the neutron flux of the fuel assembly adjacent to the fuel assembly of interest in the fuel assembly of interest A method for calculating the boundary value of the neutron diffusion equation, solving the neutron diffusion equation in the fuel assembly of interest using this boundary value, calculating the amount of change from the neutron flux in the unit cell system, and calculating the correction amount of the peaking coefficient There is. Therefore, it is possible to more accurately determine the peaking coefficient in the fuel assembly in consideration of the influence of one-sided burning with the control rod inserted and the effect of the adjacent fuel assembly. Thus, also in this embodiment, the same operational effects as those of the first to third embodiments can be obtained while improving the calculation accuracy.
[0059]
FIG. 6 is a block diagram showing a fifth embodiment of the present invention. A node output correction calculation 20 is added to the first embodiment. Generally, the node output calculator 12 solves a neutron diffusion equation and calculates a neutron flux and a heat output (density) of a unit node average, but includes a calculation error due to modeling. In the reactor core, the neutron flux distribution can be measured from the fuel assembly gap based on the local power monitor with multiple fixed neutron detectors and the reading of the movable neutron detector moving inside the tube in the axial direction. , A neutron detector 3 in the furnace is installed. The node output correction calculation unit 20 compares the neutron flux distribution calculation value calculated by the node output calculation unit 12 with the measured value of the in-core neutron detector 3 to calculate the neutron flux and the neutron flux in the fuel assembly. An output distribution correction coefficient for making the heat output distribution match the measured value is calculated. Then, using the output distribution correction coefficient, the solution of the neutron diffusion equation is corrected by the node output calculation unit 12, and the calculation error due to modeling is corrected. The above-mentioned power distribution correction coefficient corrects a calculation error caused by modeling, and has an advantage that it can be applied to the case where a future operation state of a nuclear reactor is evaluated. Thus, also in this embodiment, the same operational effects as those of the first to fourth embodiments can be obtained while improving the calculation accuracy.
[0060]
【The invention's effect】
As described above, the present invention makes it possible to accurately evaluate the thermal characteristics of a fuel assembly disposed in the core of a boiling water reactor, and to safely operate a nuclear power plant. it can. Further, since it is not necessary to determine and operate a limit value with an excessive allowance as in the related art, the fuel economy is not impaired, and the economy of the nuclear power plant can be improved.
[Brief description of the drawings]
FIG. 1 is a block diagram of a first embodiment of a method for evaluating thermal characteristics of a fuel assembly according to the present invention.
FIG. 2 is a schematic view showing an example of irregular arrangement of fuel assemblies.
FIG. 3 is a block diagram of a second embodiment of the method for evaluating thermal characteristics of a fuel assembly according to the present invention.
FIG. 4 is a block diagram of a third embodiment of the method for evaluating thermal characteristics of a fuel assembly according to the present invention.
FIG. 5 is a block diagram of a fourth embodiment of the method for evaluating thermal characteristics of a fuel assembly according to the present invention.
FIG. 6 is a block diagram showing a fifth embodiment of the method for evaluating thermal characteristics of a fuel assembly according to the present invention.
FIG. 7 is a sectional view of a core of a boiling water reactor.
FIG. 8 is a sectional view of a fuel assembly of a boiling water reactor.
FIG. 9 is a schematic view showing a regular arrangement of a central node type fuel assembly.
FIG. 10 is a schematic view showing a regular arrangement of eccentric node type fuel assemblies.
[Explanation of symbols]
1 fuel assembly
2 Control rod
3 Reactor neutron detector
4 Fuel rod
5 Water rod
6 Channel box
10 Calculation condition setting section
11 Data storage unit
12 Node output calculator
13 Node burnup calculator
14 Water gap width calculator
15 Peaking coefficient calculator
15a Unit lattice peaking coefficient calculator
15b Water gap width dependent peaking correction calculator
16 Thermal Characteristics Calculation Unit
17 Calculation result output section
18 Channel box deformation calculator
19 Peaking coefficient correction calculator
20 node output correction calculator

Claims (14)

Using the burn-up and void fraction of a fuel assembly located in the core of a boiling water reactor and the peaking coefficient calculated using the water gap width between channel boxes, the linear power density and the limit of the fuel assembly A method for evaluating thermal characteristics of a fuel assembly, comprising calculating one or both of the power ratios and evaluating the thermal integrity of the fuel assembly based on one or both of the linear power density and the critical power ratio. The method according to claim 1, wherein the peaking coefficient is calculated as a function of a burnup, a void ratio, and a width of a water gap between channel boxes. By multiplying the peaking coefficient calculated as a function of burnup and void fraction with the channel boxes arranged regularly in the same shape, the peaking correction amount for the irregular water gap width between the channel boxes is obtained. The method for evaluating thermal characteristics of a fuel assembly according to claim 1, wherein the peaking coefficient of the fuel assembly is calculated. The fuel assembly heat according to any one of claims 1 to 3, wherein the water gap width between the channel boxes is calculated using design data of individual fuel assemblies or measurement data before use. Characteristic evaluation method. The water gap width between the channel boxes is calculated by adding a deformation amount caused by neutron irradiation-induced elongation to the design data or the measurement data before use of each fuel assembly to each fuel assembly. 5. The method for evaluating thermal characteristics of a fuel assembly according to any one of 1 to 4. 6. The peaking coefficient is corrected by at least one of a neutron flux variation due to the influence of an adjacent fuel assembly and a neutron flux variation due to a single burn. Fuel assembly thermal characteristics evaluation method. 7. The heat of the fuel assembly according to claim 1, wherein a peaking coefficient of all or a specific fuel rod of the fuel assembly is calculated and a maximum value is used as a peaking coefficient of the fuel assembly. Characteristic evaluation method. 8. The method according to claim 1, wherein an R factor for calculating a limit power ratio is calculated using a result of calculating a peaking coefficient of all or a specific fuel rod of the fuel assembly. Fuel assembly thermal characteristics evaluation method. The method for evaluating thermal characteristics of a fuel assembly according to any one of claims 1 to 8, wherein a calculated value of a node power distribution of the fuel assembly is corrected using a signal of a neutron detector in the reactor. . A calculation condition setting unit that sets the operating state of the boiling water reactor to be evaluated and stores the set operation state in a data storage unit, and an operation state set in the calculation condition setting unit. A node output calculator for calculating a unit node average neutron flux, a heat output, and a void fraction distribution with a region obtained by dividing the fuel assembly into a plurality in the axial direction as a unit node, and storing the calculated data in the data storage unit; Using the average heat output and the void fraction distribution, calculate the unit node burnup and store it in the data storage unit using the node burnup calculator and the design data of each fuel assembly or the measurement data before use. Calculates the water gap width between channel boxes and stores it in the data storage unit, and uses the unit node burnup, void fraction distribution, and water gap width to collect fuel. The peaking coefficient of the fuel assembly is calculated using the peaking coefficient calculation unit that stores the peaking coefficient in the data storage unit, the unit node average heat output, the void fraction distribution, and the peaking coefficient. And a calculation result output unit for summarizing the calculation results and outputting information and detailed information of the most severe fuel assembly in the reactor, characterized in that: Evaluation device. A unit node peaking coefficient calculation unit that calculates the peaking coefficient as a function of the burnup and void fraction in a state where the channel boxes are regularly arranged in the same shape, and a case where the water gap width between the channel boxes becomes irregular. The thermal characteristic evaluation apparatus for a fuel assembly according to claim 10, further comprising a peaking coefficient calculator having a water gap width dependent peaking correction calculator for calculating a peaking correction amount. The fuel assembly thermal characteristic evaluation device according to claim 10 or 11, further comprising a channel box deformation amount calculation unit configured to calculate a deformation amount caused by elongation due to neutron irradiation of the channel box. 11. A peaking coefficient correction calculator for correcting a peaking coefficient by at least one of a change amount of a neutron flux due to an influence of an adjacent fuel assembly and a change amount of a neutron flux due to an influence of one-sided burning. 13. The fuel assembly thermal characteristic evaluation device according to any one of claims 12 to 12. The fuel according to any one of claims 10 to 13, further comprising: a node output correction calculation unit that corrects a calculated value of a node output distribution of the fuel assembly by using a signal of the in-core neutron detector. Aggregate thermal property evaluation device.

Images