JP4526781B2 - Method and apparatus for evaluating fuel assembly thermal characteristics - Google Patents

Description

[0001]
BACKGROUND 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. In the core, 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. Each fuel assembly 1 is configured as shown in FIG. 8, and 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 soundness of the fuel assembly 1 during operation of the boiling water reactor, the generated heat output per unit length of the fuel rod 2 constituting the fuel assembly 1 (hereinafter referred to as “linear power density”) Must be kept below the limit value. In addition, even in the case of abnormal transient changes during operation, the limit output and normal operation should be avoided so that the fuel rod damage caused by insufficient cooling does not occur (hereinafter referred to as “limit output”). It is necessary that the ratio (hereinafter referred to as “limit output ratio”) to the generated heat output of the fuel assembly 1 at the time be equal to or greater than the limit value.
[0004]
In the prior art, in the monitoring or prediction of the core, the line power density and the limit power ratio are generally calculated and evaluated 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 by approximating that the fuel composition is homogeneous within the unit nodes, a small group or a modified group of neutrons are obtained. Solve the diffusion equation and calculate the unit node average neutron flux and thermal power (density). Next, by multiplying the peaking coefficient ("local peaking coefficient") of the fuel rod that produces the most heat output in the fuel assembly by the thermal output (density) of the unit node average, the linear output density of the fuel rod that is the maximum output is obtained. calculate. It is evaluated that the thermal integrity of the fuel assembly 1 can be ensured if the linear power density obtained as described above is controlled to be equal to or less than a predetermined limit value.
[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. The empirical formula is organized using a factor related to the peaking coefficient called the R factor together with the coolant flow rate, pressure, and fuel assembly shape factor data. Therefore, the coolant flow rate, pressure, and shape of each fuel assembly 1 are arranged. The limit output is determined by the factor and the R factor. The limit power ratio is calculated as the ratio between the limit output of each fuel assembly 1 and the actual generated heat output. It is evaluated that the thermal integrity of the fuel assembly 1 can be ensured if the limit output ratio obtained as described above is controlled to be equal to or greater than a predetermined limit value.
[0007]
The above calculation requires a local peaking coefficient for calculating the linear power density and an R factor for calculating the limit power ratio as the peaking coefficient for each fuel assembly 1. The local peaking coefficient and the R factor are calculated by a unit fuel assembly core calculation code, assuming that fuel assemblies 1 of the same shape are regularly arranged, and the burnup and voids of the fuel assembly 1 are calculated. Generally given as a function of rate. FIG. 9 is a schematic diagram showing a regular arrangement of the central lattice type fuel assemblies 1, and the water gap width (W) between the channel boxes of each fuel assembly is all equal. FIG. 10 is a schematic diagram showing a regular arrangement of the eccentric lattice type fuel assemblies 1, and the width (Ww) on the control rod insertion side of the water gap width between the channel boxes of each fuel assembly is controlled. Although it is a lattice type wider than the width (Wn) on the non-insertion side of the rod, the values of Ww are equal and the values of Wn are equal for all the fuel assemblies.
[0008]
The peaking coefficient used in the conventional technique is calculated on the assumption that the fuel assemblies having the same shape are regularly arranged as shown in FIG. 9 or FIG. ing.
[0009]
[Patent Literature]
Japanese Patent Application No. 2001-360665
[0010]
[Problems to be solved by the invention]
However, in future fuel assemblies, it has been considered to use channel boxes having different shapes. Normally, not all fuel assemblies are replaced during a single periodic inspection, but instead of 1/5 to 1/3, there are channel boxes with different shapes 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 drawback that the thermal integrity of the fuel assemblies cannot be accurately evaluated. Also, with the conventional evaluation method, even when using a channel box with the same shape, the manufacturing tolerance of the channel box and deformation during use cannot be taken into consideration, so a limit value with an excessive margin corresponding to the evaluation error is set. It is necessary to determine and operate it, and this will impair the fuel economy.
[0011]
As described above, the conventional evaluation method has a drawback that the thermal soundness of the fuel assembly cannot be accurately evaluated when channel boxes having different shapes are used. Even when channel boxes having the same shape are used, it is necessary to determine and operate a limit value with an excessive margin, which impairs fuel economy.
[0012]
In view of these points, the present invention is to obtain a fuel assembly thermal characteristic evaluation method and apparatus capable of evaluating in detail the thermal integrity of a fuel assembly disposed in the core of a boiling water reactor. Objective.
[0013]
[Means for Solving the Problems]
  The invention according to claim 1
  The fuel assembly arranged in the core of the boiling water reactor is divided into a plurality of unit nodes in the axial direction and adjacent to the four sides of the burn-up rate and void ratio of each unit node and the channel box of the target unit node. The line output density of the unit node calculated using the peaking coefficient of each unit node calculated using the four water gap widths obtained as the intervals between the four channel boxes and the average heat output of each unit node; ,
  Fuel assembly limit output ratio calculated using the peaking coefficient, the fuel assembly limit output, and the average heat output of the fuel assembly;
One or both of these are calculated, and the thermal integrity of the fuel assembly is evaluated by 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 burnup, void ratio, and water gap width between channel boxes.
[0015]
  The invention according to claim 3 is the invention according to claim 1, in which the channel box is regularly arranged in the same shape with respect to the fuel assembly arranged in the core of the boiling water reactor.Unit nodeThe peaking coefficient calculated as a function of burnup and void fractionWith four adjacent fuel assembliesCalculate the peaking coefficient of the fuel assembly by multiplying the peaking correction amount due to irregular water gap width between channel boxes, and evaluate one or both of the fuel assembly linear power density and the critical power ratio It is characterized by.
[0016]
  The invention according to claim 4 is the invention according to any one of claims 1 to 3, wherein the fuel assembly disposed in the core of the boiling water reactor isWith four adjacent fuel assembliesThe water gap width between channel boxes is calculated using design data of individual fuel assemblies or measurement data before use.
[0017]
The invention according to claim 5 is the invention according to any one of claims 1 to 4, wherein the water gap width between the channel boxes is set individually for the fuel assembly disposed in the core of the boiling water reactor. The calculation is performed by adding the amount of deformation caused by the neutron irradiation of the channel box 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, the neutron flux of the fuel assembly disposed in the core of the boiling water reactor is influenced by the adjacent fuel assembly. The peaking coefficient is corrected by at least one of the change amount and the change amount of the neutral bundle due to the influence of one burn.
[0019]
The invention according to claim 7 is the invention according to any one of claims 1 to 6, wherein the peaking coefficient of all or specific fuel rods is calculated for the fuel assemblies arranged in the core of the boiling water reactor. The maximum value is 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 coefficients of all or specific fuel rods are calculated for the fuel assemblies arranged in the core of the boiling water reactor. The result is used to calculate the R factor for calculating the limit power ratio.
[0021]
The invention according to claim 9 is the invention according to any one of claims 1 to 8, wherein the fuel assembly disposed in the core of the boiling water reactor is utilized by using a signal of the in-core neutron detector. The calculation value of the node output distribution of the fuel assembly is corrected.
[0022]
  The invention according to claim 10 sets the operation state of the boiling water reactor to be evaluated, and stores the calculation condition setting unit stored in the data storage unit and the operation state set by the calculation condition setting unit. A node output calculation unit that calculates a unit node average neutron flux, heat output, and void ratio distribution with a unit node as a region obtained by dividing the fuel assembly into a plurality of axial directions, and stores it in the data storage unit; Using the node average heat output and void fraction distribution, the unit node burnup is calculated and stored in the data storage unit, and the individual fuel assembliesDesign of fuel assemblies that use design data or pre-use measurement data to assess thermal healthChannel boxThe four distances calculated as the distances between the four sides of the fuel cell and the channel boxes of the four adjacent fuel assembliesThe water gap width calculation unit that calculates the water gap width and stores it in the data storage unit, unit node burnup and void ratio distributionThe fourThe fuel assembly peaking coefficient is calculated using the water gap width, and the peaking coefficient calculation unit stored in the data storage unit, the unit node average heat output, void fraction distribution, and peaking coefficient are used to calculate the fuel assembly line output. It has a thermal characteristic calculation unit that calculates one or both of density and limit power ratio, and a calculation result output unit that aggregates the calculation results and outputs the most severe fuel assembly information and detailed information in the reactor And
[0023]
  The invention according to claim 11 is the invention according to claim 10, wherein the peaking coefficient in a state where the channel boxes are regularly arranged in the same shape is obtained.Unit nodeA unit node peaking coefficient calculator to calculate as a function of burnup and void fraction;The fourIt has a peaking coefficient calculation unit having a water gap width dependent peaking correction calculation unit for calculating a peaking correction amount due to irregularity of the water gap width.
[0024]
The invention according to claim 12 is the invention according to claim 10 or 11, further comprising a channel box deformation amount calculation unit for calculating a deformation amount due to elongation of the channel box due to neutron irradiation.
[0025]
The invention according to claim 13 is the invention according to any one of claims 10 to 12, wherein 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 neutral flux due to an influence of a single burn. It has a peaking coefficient correction calculator for correcting the peaking coefficient.
[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 from a reactor neutron detector It has the part.
[0027]
DETAILED DESCRIPTION OF THE INVENTION
Embodiments of the present invention will be described below. 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 in the calculation condition setting unit 10, the set operation condition is stored in the data storage unit 11. . The data storage unit 11 includes a unit node average neutron flux calculated by the node output calculation unit 12, a heat output (density) and a void rate distribution, a unit node burnup calculated by the node burnup calculation unit 13, a water The water gap width calculated by the gap width calculation unit 14 and the peaking coefficient calculated by the peaking coefficient calculation unit 15 are respectively input and stored.
[0029]
That is, the node output calculation unit 12 solves the neutron diffusion equation for the operating state set by the calculation condition setting unit 10, and calculates the unit node average neutron flux, heat output (density), and void ratio distribution, The unit node average neutron flux, heat output (density), and void fraction distribution are stored in the data storage unit 11. The node burnup distribution calculation unit 13 calculates a 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. The 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. In the output peaking coefficient calculation unit 15, the unit node burnup calculated by the node burnup calculation unit 13, the void ratio distribution calculated by the node output calculation unit 12, and the water gap width calculated by the water gap width calculation unit 14 are calculated. The peaking coefficient of the fuel assembly is calculated and stored in the data storage unit 11.
[0030]
The thermal characteristic calculation unit 16 uses the unit node average thermal output calculated by the node output calculation unit 12 and the peaking coefficient calculated by the peaking coefficient calculation unit 15 to determine the line output density and the limit output ratio of the fuel assembly. The calculation result output unit 17 calculates the calculation results and outputs the strictest fuel assembly information and detailed information in the nuclear reactor.
[0031]
The core of the boiling water reactor to be evaluated is configured as shown in FIG. 7, as in the prior art. In the core, 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. 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 with 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 representative example, in a 1.1 million class boiling water reactor, If there are 764 fuel assemblies 1 and the number of divisions in the axial direction is 24, there will be 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. ), Or modified group 1 (modified when the neutron group consists of only one group of fast neutrons), coupled with nuclear thermohydraulic calculation, unit node average neutron flux, heat Calculate power (density) and void fraction distribution.
[0034]
The node burnup calculation unit 13 obtains a unit node burnup increment by multiplying the unit node average heat output by the time increment from the starting point state, and adds the unit node burnup increment to the unit node burnup 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 performs the calculation for the 4 of the channel box 6 of the fuel assembly 1 (the central fuel assembly in FIG. 2) to be calculated with respect to the irregular arrangement of the fuel assemblies 1 as shown in FIG. For each surface, the water gap width (Wa, Wb, Wc, Wd) is calculated as the distance from the channel box 6 of the adjacent fuel assembly 1. As the geometrical shape of the channel box 6 used for calculating the water gap width, there are a method using design data on each channel box shape and installation position, and a method using measurement data before use. Since the channel box 6 has a height of about 4 m, there may be a bend in the middle. Therefore, when the bend data at the axial position exists as the measurement data before use, the channel box 6 Using measured data as a bend, water gap width is axialForIt can be calculated in detail for each position.
[0036]
The peaking coefficient calculation unit 15 calculates the peaking coefficient and R factor by the following equations (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: Fuel assembly burnup
Wa, Wb, Wc, Wd: Water gap width
fl: function for calculating the local peaking coefficient
fr: function for calculating the R factor
[0037]
There are various methods for creating a function for calculating the local peaking coefficient used in the above formula and a function for calculating the R factor. In this embodiment, the function is calculated in advance using a unit fuel assembly calculation code. Here are some examples:
[0038]
The local peaking coefficient in the unit node is an index unique to the unit node that is generally distinguished by the geometric shape of the fuel assembly in the unit node of interest, the concentration of each fuel rod, the distribution of flammable poisons such as gadolinia, etc. Organized by (fuel type). For each fuel type, individual water gap widths are in multiple states (Wa1, Wa2, Wa3 ...), (Wb1, Wb2, Wb3 ...), (Wc1, Wc2, Wc3 ...), (Wd1, Wd2, Wd3 ...) Assuming that the local peaking coefficient is calculated in advance using the unit fuel assembly calculation code for each of the burnups E1, E2, E3, and the void ratios V1, V2, V3, and so on. It is stored in the data storage unit 11 as a number table or a number table format of coefficients of the fitting equation. As a fitting expression, there are a method of storing a function of a polynomial of a void rate for each burnup Ei (i = 1, 2,...) And a method of storing a function of a polynomial relating to a burnup E and a void rate V. . In addition, the number table and the number table of the coefficients of the fitting equation are obtained in two ways for each of the cases where the control rod 2 is not included in the unit node of interest, and each is stored.
[0039]
The fuel assembly R-factor is organized by a fuel assembly-specific index (bundle type) that is distinguished by the geometric shape of the fuel assembly, the concentration of each fuel rod, the distribution of flammable poisons such as gadolinia, etc. The For each bundle type, individual water gap widths are in multiple states (Wa1, Wa2, Wa3 ...), (Wb1, Wb2, Wb3 ...), (Wc1, Wc2, Wc3 ...), (Wd1, Wd2, Wd3 ...) Assuming that the R factor is calculated in advance using the unit fuel assembly calculation code for each of the burnups E1, E2, E3... Is stored in the data storage unit 11. The number table and the number table of the coefficients of the fitting formula are stored as a number table according to the ratio at which the control rod 2 is inserted in the fuel assembly of interest or a number table of the coefficients of the fitting formula.
[0040]
There are various other methods for handling the water gap width in the function for calculating the local peaking coefficient and the function for calculating the R factor. For example, instead of using the individual water gap widths as they are, there is a method of organizing by deviation from design data when the channel boxes are regularly arranged in the same shape. There are also a method of processing the absolute value of the deviation with the maximum value of the four surfaces and a method of processing with the average value of the four surfaces of the absolute value of the deviation. Furthermore, considering the symmetry of the design of the fuel assembly, the processing is performed with the maximum value of Wa and Wb and the maximum value of Wc and Wd, or 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 line output density for each unit node by the following formula using the unit node average heat output and the local peaking coefficient.
LHGR = α * PN * LPF ……………… (3)
here,
LHGR: Unit node line power density
PN: Average thermal 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 the fuel assembly
fc: Empirical formula for calculating the limit output
FB: Fuel assembly coolant flow rate
RF: R factor of fuel assembly
PR: Reactor pressure
β: fuel assembly form factor
[0043]
The calculation result output unit 17 tabulates the line power density and the limit power ratio, which are outputs from the thermal characteristic calculation unit 16, and outputs the most severe values in the core and information on individual fuel assemblies.
[0044]
As described above, the present invention calculates the fuel assembly linear power density and the critical power ratio using the peaking coefficient calculated as a function of the burnup rate, the void ratio, and the water gap width between the channel boxes. For example, the thermal soundness of the fuel assembly can be accurately evaluated by comparing the value with the limit value.
[0045]
FIG. 3 is a block diagram showing the second embodiment of the present invention. The peaking coefficient calculation unit 15 of the first embodiment is replaced with a unit cell peaking calculation unit 15a and a water gap width dependent peaking correction calculation unit. It is separated into 15b. The unit lattice peaking coefficient calculation unit 15a calculates the peaking coefficient as a function of the burnup rate and the void ratio in a state where the channel boxes are regularly arranged in the same shape (unit node). Further, the water gap width-dependent peaking correction calculation unit 15b calculates a peaking correction amount due to an irregular water gap width between channel boxes. That is, the peaking coefficient calculation unit 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: Fuel assembly burnup
Wa, Wb, Wc, Wd: Water gap width
LIJ: Location of the fuel rod that generates the local peaking coefficient
RIJ: Location of R-factor generating fuel rod
fld: function for calculating the local peaking coefficient of the unit cell
flg: a function for correcting the local peaking coefficient depending on the water gap width
number
frd: function for calculating the R factor of the unit cell
frg: a function for correcting the R factor depending on the water gap width
[0046]
The local peaking coefficient of the unit cell is generally a unit node-specific index (identified by the geometric shape of the fuel assembly in the unit node of interest, the concentration of each fuel rod, the distribution of flammable poisons such as gadolinia, etc.) “Fuel type”). For each fuel type, for each of the burnups E1, E2, E3 ..., for each of the void rates V1, V2, V3 ..., the local peaking coefficient is calculated in advance using a unit fuel assembly calculation code, The values are 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 a fitting expression, there are a method of storing a function of a polynomial of a void rate for each burnup Ei (i = 1, 2,...) And a method of storing a function of a polynomial relating to a burnup E and a void rate V. . In addition, the number table and the number table of the coefficients of the fitting equation are obtained in two ways for each of the cases where the control rod 2 is not included in the unit node of interest, and each is stored.
[0047]
The function that corrects the local peaking coefficient depending on the water gap width is 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 using a unit fuel assembly calculation code for fuel assemblies with uniform enrichment, and the channel box is the same shape and regular And a sensitivity function for each fuel rod position is determined in advance. The above sensitivity functions are calculated for a plurality of void ratios V1, V2, V3... And stored in the data storage unit 11 as a numerical table or a numerical table format of coefficients of a fitting equation. In addition, the number table and the number table of the coefficients of the fitting equation are obtained in two ways for each of the cases where the control rod 2 is not included in the unit node of interest, and each is stored.
[0048]
The R factor of the unit cell is organized by an index (bundle type) unique to the fuel assembly, which is distinguished by the geometric shape of the fuel assembly, the concentration of each fuel rod, the distribution of flammable poisons such as gadolinia, etc. . For each bundle type, an R factor is calculated in advance using a unit fuel assembly calculation code for each of the burnups E1, E2, E3... The data is stored in the data storage unit 11 as a format. The number table and the number table of the coefficients of the fitting formula are stored as a number table according to the ratio at which the control rod 2 is inserted in the fuel assembly of interest or a number table of the coefficients of the fitting formula.
[0049]
The function for correcting the R-factor depending on the water gap width is such that the water gap width is in a plurality of states (Wa1, Wa2, Wa3 ...), (Wb1, Wb2, Wb3 ...), (Wc1, Wc2, Wc3 ...), ( Assuming Wd1, Wd2, Wd3, etc.), the R factor is calculated using the unit fuel assembly calculation code for fuel assemblies with uniform enrichment, and the channel boxes are regularly arranged in the same shape. The sensitivity function for each fuel rod position is determined in advance in comparison with the R factor in the state (unit cell). The sensitivity function is stored in the data storage unit 11 as a number table according to the ratio of the control rods 2 inserted into the fuel assembly of interest or a number table of coefficients of the fitting equation.
[0050]
There are various other methods for handling the water gap width 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. For example, instead of using the individual water gap widths as they are, there is a method of organizing by deviation from design data when the channel boxes are regularly arranged in the same shape. There are also a method of processing the absolute value of the deviation with the maximum value of the four surfaces and a method of processing with the average value of the four surfaces of the absolute value of the deviation. Furthermore, considering the contrast of the fuel assembly design, the processing is performed with the maximum value of Wa and Wb and the maximum value of Wc and Wd, or 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 specific fuel rods in the unit cell fuel assembly, and correction is made for irregular water gap width between channel boxes. As a result, there is a method of determining the local peaking coefficient and R factor of the fuel assembly by obtaining the maximum value in the fuel assembly.
[0052]
Further, the peaking coefficient is stored in the data storage unit 11 for all or specific fuel rods in the unit cell fuel assembly, and the correction result for irregularity of the water gap width between the channel boxes is obtained. There is also a method in which the R factor of the fuel assembly is determined by calculating the R factor using it and obtaining the maximum value in the fuel assembly.
[0053]
Therefore, also in this embodiment, the same operational effects as in the first embodiment can be achieved while reducing the necessary data amount.
[0054]
FIG. 4 is a block diagram showing a third embodiment of the present invention, and a channel box deformation amount calculation unit 18 is added to the first embodiment. The channel box deformation amount calculation unit 18 calculates a bend 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 bend amount of the channel box calculated by the channel box deformation amount calculation unit 18 to the design data of each fuel assembly or the measurement data before use, and thereby the water gap between the channel boxes. Calculate the width.
[0055]
The channel box deformation amount calculation unit 18 calculates the elongation amounts in the longitudinal direction of the four surfaces of the channel box using the neutron irradiation amounts of the four surfaces of the channel box stored in the data storage unit 11, and the difference in elongation therebetween. Calculate the amount of bending in the channel box. For this calculation, the neutron flux of the four faces of the channel box is calculated and stored in the data storage unit 11 when the node output calculation unit 12 calculates the node output. In addition, when the node burnup is calculated by the node burnup calculation unit 13, the neutron irradiation amounts of 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 neutron irradiation on the four faces of the channel box from the difference in burnup increments of adjacent fuel assemblies. Obtain the gradient of the amount, calculate the difference in elongation of the channel box from the gradient, and calculate the amount of bending of the channel box.
[0057]
In this way, in this embodiment, the peaking coefficient in the fuel assembly can be obtained more accurately using the water gap width considering the deformation of the channel box. In this embodiment, the same operational effects as those of the first and second embodiments are obtained while improving the calculation accuracy.
[0058]
FIG. 5 is a block diagram showing a fourth embodiment of the present invention, and a peaking coefficient correction calculation unit 19 is added to the first embodiment. In the first embodiment, when the peaking coefficient calculation unit 15 calculates the peaking coefficient, a numerical table that is a calculation result of the unit cell system or a numerical table of coefficients of the fitting equation is used. However, the neutron distribution in the fuel assembly placed in the nuclear reactor is affected by the single burn with the control rod 2 inserted and by the adjacent fuel assembly. There is a slight difference. Therefore, in order to more accurately determine the peaking coefficient in the fuel assembly placed in the nuclear reactor, the peaking coefficient of the unit cell system is influenced by the influence of single burn with the control rods inserted and the adjacent fuel assembly. It is necessary to consider the influence of the body. In the fourth embodiment, a peaking coefficient correction calculation unit 19 is added to perform the above correction. As an example of the calculation method of the peaking coefficient correction calculation unit 19, the influence of the single burn in the state where the control rod has been inserted is the amount of change in the peaking coefficient proportional to the control rod insertion period. There is a method of correcting with a table. In addition, as a method of considering the influence of adjacent fuel assemblies, 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 A method for calculating the correction value of the peaking coefficient by 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 of the unit cell system There is. Therefore, it is possible to more accurately determine the peaking coefficient in the fuel assembly in consideration of the influence of the single burn 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 to third embodiments are obtained while improving the calculation accuracy.
[0059]
FIG. 6 is a block diagram showing the fifth embodiment of the present invention, and a node output correction calculation 20 is added to the first embodiment. In general, the node output calculation unit 12 solves the neutron diffusion equation and calculates the average neutron flux and heat output (density) of the unit nodes, 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 with multiple fixed neutron detectors and the reading of the movable neutron detector moving inside the tube in the axial direction. In-furnace neutron detector 3 is installed. The node output correction calculation unit 20 compares the calculated value of the neutron flux distribution in the gap between the fuel assemblies calculated by the node output calculation unit 12 with the measurement value of the in-core neutron detector 3 to determine the neutron flux in the fuel assembly. An output distribution correction coefficient is calculated so that the thermal output distribution matches the measured value. Then, using the output distribution correction coefficient, the node output calculation unit 12 corrects the solution of the neutron diffusion equation, and the calculation error due to modeling is corrected. The above power distribution correction coefficient corrects a calculation error caused by modeling, and has an advantage that can be applied when evaluating the future operating state of the nuclear reactor. Therefore, this embodiment also has the same operational effects as the first to fourth embodiments while improving the calculation accuracy.
[0060]
【The invention's effect】
As described above, the present invention can accurately evaluate the thermal characteristics of the fuel assembly arranged in the core of the boiling water reactor, and can safely operate the nuclear power plant. it can. Further, since it is not necessary to set and operate a limit value with an excessive margin as in the prior art, it is possible to improve the economic efficiency of the nuclear power plant without impairing the economic efficiency of the fuel.
[Brief description of the drawings]
FIG. 1 is a block configuration diagram of a first embodiment of a fuel assembly thermal characteristic evaluation method of the present invention.
FIG. 2 is a schematic diagram showing an example of irregular arrangement of fuel assemblies.
FIG. 3 is a block configuration diagram of a second embodiment of the fuel assembly thermal characteristic evaluation method of the present invention.
FIG. 4 is a block diagram of a third embodiment of the fuel assembly thermal characteristic evaluation method of the present invention.
FIG. 5 is a block diagram of a fourth embodiment of the fuel assembly thermal characteristic evaluation method of the present invention.
FIG. 6 is a block diagram of a fifth embodiment of the fuel assembly thermal characteristic evaluation method of the present invention.
FIG. 7 is a cross-sectional view of the core of a boiling water reactor.
FIG. 8 is a cross-sectional view of a fuel assembly of a boiling water reactor.
FIG. 9 is a schematic diagram showing a regular arrangement of a central node type fuel assembly.
FIG. 10 is a schematic diagram showing a regular arrangement of eccentric node type fuel assemblies.
[Explanation of symbols]
1 Fuel assembly
2 Control rod
3 In-core neutron detector
4 Fuel rod
5 Water rod
6 Channel box
10 Calculation condition setting section
11 Data storage
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 calculator
17 Calculation result output section
18 Channel box deformation calculator
19 Peaking coefficient correction calculator
20 node output correction calculator

Claims (14)

The fuel assembly arranged in the core of the boiling water reactor is divided into a plurality of unit nodes in the axial direction and adjacent to the four sides of the burn-up rate and void ratio of each unit node and the channel box of the target unit node. The line output density of the unit node calculated using the peaking coefficient of each unit node calculated using the four water gap widths obtained as the intervals between the four channel boxes and the average heat output of each unit node; ,
Fuel assembly limit output ratio calculated using the peaking coefficient, the fuel assembly limit output, and the average heat output of the fuel assembly;
A fuel assembly thermal characteristic evaluation method characterized by calculating one or both of the above and evaluating the thermal integrity of the fuel assembly based on one or both of the linear power density and the limit power ratio.   2. The fuel assembly thermal characteristic evaluation method according to claim 1, wherein the peaking coefficient is calculated as a function of burnup, void ratio, and water gap width between channel boxes. The peak gap coefficient calculated as a function of the burnup and void fraction of the unit nodes with the channel boxes regularly arranged in the same shape does not show the water gap width between the channel boxes of the four adjacent fuel assemblies. 2. The fuel assembly thermal characteristic evaluation method according to claim 1, wherein the peaking coefficient of the fuel assembly is calculated by multiplying the peaking correction amount for becoming a rule. The water gap width between the channel boxes of four adjacent fuel assemblies is calculated using design data of individual fuel assemblies or measurement data before use. The fuel assembly thermal characteristic evaluation method according to any one of the above.   The water gap width between channel boxes is calculated by adding the amount of deformation caused by elongation due to neutron irradiation of the channel box to the design data of individual fuel assemblies or measurement data before use. 5. The fuel assembly thermal characteristic evaluation method according to any one of 1 to 4.   6. The peaking coefficient is corrected 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 neutral flux due to an influence of a single burn. For evaluating the fuel assembly thermal characteristics.   The fuel assembly heat according to any one of claims 1 to 6, wherein the peaking coefficient of the fuel assembly is calculated by calculating a peaking coefficient of the whole fuel assembly or a specific fuel rod. Characterization method.   8. The R factor for calculating the limit power ratio is calculated using the result of calculating the peaking coefficient of all of the fuel assemblies or a specific fuel rod. 8. Fuel assembly thermal characteristics evaluation method.   9. The fuel assembly thermal characteristic evaluation method according to claim 1, wherein a calculated value of the node output distribution of the fuel assembly is corrected using a signal from the in-core neutron detector. . Set the operation state of the boiling water reactor to be evaluated, and store the set operation state in the data storage unit, and the operation state set by the calculation condition setting unit A node output calculation unit for calculating a unit node average neutron flux, heat output and void ratio distribution with a unit node as a region obtained by dividing the fuel assembly into a plurality of axial directions, and storing it in the data storage unit; Using the average heat output and void fraction distribution, the unit node burnup is calculated and stored in the data storage unit, and the design data of each fuel assembly or measurement data before use is used. calculate the four water gap width determined as a distance between the four channel box of a fuel assembly adjacent to the four sides of the channel box of a fuel assembly for evaluating the thermal health, data And the water gap width calculation unit to be stored in憶部, the units node burnup and void fraction distribution by using four water gap width to calculate the peaking factor of the fuel assembly, peaking coefficient calculation unit be stored in the data storage unit A thermal characteristic calculator that calculates one or both of the linear power density and the critical power ratio of the fuel assembly using the unit node average thermal power, void fraction distribution, and peaking coefficient, A fuel assembly thermal characteristic evaluation apparatus comprising a calculation result output unit for outputting the most severe fuel assembly information and detailed information. The unit node peaking coefficient calculation unit for calculating the peaking coefficient in a state where the channel boxes are regularly arranged in the same shape as a function of the burnup and void ratio of the unit node, and the four water gap widths are irregular. 11. The fuel assembly thermal characteristic evaluation apparatus according to claim 10, further comprising a peaking coefficient calculation unit having a water gap width-dependent peaking correction calculation unit for calculating a peaking correction amount for saving.   The fuel assembly thermal characteristic evaluation apparatus according to claim 10 or 11, further comprising a channel box deformation amount calculation unit for calculating a deformation amount due to elongation of the channel box due to neutron irradiation.   11. A peaking coefficient correction calculation unit 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 neutral flux due to an influence of a single burn. The fuel assembly thermal characteristic evaluation device according to any one of 1 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 the node output distribution of the fuel assembly using a signal from the in-core neutron detector. Aggregate thermal property evaluation system.

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