JP4966900B2 - Method and apparatus for calculating core performance of boiling water reactor - Google Patents

Description

  The present invention relates to a core performance calculation apparatus, and more particularly to a core performance calculation method suitable for application to a boiling water reactor loaded with a fuel assembly having a spectrum shift rod in which the internal water level changes due to a change in core flow rate. And device.

  Conventionally, in boiling water reactors, core characteristics such as power distribution and reactivity in the core are monitored, and changes in core characteristics when the control rod position and core flow rate are changed are predicted to make the reactor safe. A core performance calculator is provided for control. The core performance calculation device calculates the nuclear constant of the reactor core by thermal hydraulic calculation, then calculates the three-dimensional neutron flux distribution in the core by nuclear calculation such as neutron transport calculation or neutron diffusion calculation, and calculates the core characteristics .

  In this calculation, the core is divided into three-dimensional regions and nuclear constants reflecting the material composition in each region are calculated. The nuclear constant is related to the nuclear reaction by neutrons and changes with the operation of the reactor. Therefore, a detailed assembly calculation is performed assuming an infinite lattice system in which the same fuel assemblies are arranged indefinitely, and the neutron distribution of the fuel assemblies and the composition of the nuclear fuel material are tracked. The characteristics of the fuel assembly are being sought. The calculation method is to perform assembly calculation in several cases with different moderator densities for one type of fuel assembly, and calculate the assembly nuclear constant including the nuclear reaction cross section of the fuel assembly average and neutron infinite multiplication factor. Sort by values such as burn-up, moderator density, and burner average moderator density.

  In an actual core, the moderator density and the burnup value differ depending on the position in the core. For this reason, for example, as described in Patent Document 1, in the core calculation for obtaining the neutron distribution over the entire core and the reactivity of the entire core, the burnup, moderator density, combustion obtained separately from the results of the aggregate calculation The aggregate nuclear constant corresponding to the state in each region in the furnace is calculated by interpolating with values such as the average moderator density.

  On the other hand, from the viewpoint of effective use of nuclear fuel material, a fuel assembly having a spectrum shift rod (hereinafter referred to as SSR) as shown in Patent Document 2 has been proposed. The SSR is an inverted U-shaped tube, and is a water rod in which the boundary between the water vapor region and the water region (hereinafter referred to as the water level in the SSR) formed inside the SSR changes as the core flow rate changes. The fuel assembly equipped with this SSR can control the moderator density in the core over a wide range by changing the core flow rate, and can promote the conversion from uranium 238 to plutonium 239 and improve the reactivity at the end of combustion. Effective use of nuclear fuel materials can be achieved. Since the water level in the SSR changes as the core flow rate changes, it is important to perform the core calculation in consideration of the water level in the SSR in order to accurately calculate the core performance.

  In the reactor performance calculation apparatus described in Patent Document 3, in a nuclear reactor loaded with a fuel assembly having SSR, the nuclear level in each region in the core is taken into account in consideration of the water level in the SSR that changes according to the core flow rate. By obtaining the number, core characteristics such as power distribution and reactivity in the core are accurately calculated.

JP-A-4-320996 JP-A 63-73187 JP 7-92289 A

In the core performance calculation method described in Patent Document 3, the pressure loss ΔP _LTP when the cooling water supplied to the core passes through the lower tie _plate is equal to the total pressure loss in the SSR, and at the SSR inlet and outlet. Assuming that the pressure loss is small, the water level Z in the SSR is calculated from the equation (1) using the density ρ _{f of} water and the gravitational acceleration g, and the nuclear constant is calculated based on the presence or absence of water in the SSR. .
Z = ΔP _LTP / ρ _f g (1)
This water level calculation method is simple and excellent from the viewpoint of high-speed calculation important for core performance calculation. However, if the cooling water flow rate supplied to each fuel assembly or the pressure loss at the lower tie plate is different from the actual cooling water flow rate or the lower tie plate pressure loss, the calculated water level in the SSR and the actual There is a difference in the water level in the SSR, which may cause errors in the core performance calculation.

  An object of the present invention is to provide a core performance calculation method and apparatus for further improving the calculation accuracy of core characteristics in a boiling water reactor loaded with a fuel assembly having SSR.

  The present invention includes a plurality of fuel assemblies loaded on a core, a spectrum shift rod that is provided in the fuel assemblies and whose internal water level changes in accordance with the cooling water flow rate of the core, and is axially movable in the core. For the boiling water reactor with a neutron detector installed in the reactor, the core performance calculation method for the boiling water reactor, which calculates the core characteristics including the power distribution in the reactor by the nuclear thermal hydraulic coupling calculation, The core axial power distribution calculated based on the neutron flux distribution in the core and the axial power distribution calculated by the nuclear thermal hydraulic coupling calculation are compared, and the actual water level in the spectrum shift rod and the core performance are calculated. The fuel assembly in which the difference in the set water level has occurred is identified, and the water level in the already set spectrum shift rod is adjusted so as to match the actual water level in the spectrum shift rod. Calculated and correcting.

  Further, in the core performance calculation method for a boiling water reactor, the nuclear thermal hydraulic coupling calculation distributes a cooling water flow rate for each channel of the fuel assembly, and calculates a void ratio distribution and a spectrum shift in the fuel assembly. It is characterized by calculating the core constant and neutron distribution of the core based on the water level calculation in the rod, and calculating the core characteristics including the power distribution in the reactor.

  Further, in the core performance calculation method of the boiling water reactor, the fuel in which the difference between the actual water level in the spectrum shift rod caused by the difference in the cooling water flow rate and the water level in the spectrum shift rod set during the core performance calculation has occurred The aggregate is specified, and the water level in the already set spectrum shift rod is recalculated and corrected so as to match the water level in the actual spectrum shift rod.

  Also, in the core performance calculation method for boiling water reactors, the actual water level in the spectral shift rod is calculated based on the difference in reactivity to the water level change in the spectral shift rod caused by the difference in the nuclear characteristics of the fuel assembly. The fuel assembly in which the difference in the water level in the spectrum shift rod set at the time of calculating the core performance is identified, and the water level in the spectrum shift rod that has already been set is reset to match the water level in the actual spectrum shift rod. It is characterized by calculating and correcting.

  Further, in the core performance calculation method for a boiling water reactor, the void ratio in the spectral shift rod corresponding to the cooling water flow rate of the fuel assembly calculated and stored in advance when the water level in the spectral shift rod is recalculated. The table information regarding is used.

  Furthermore, a plurality of fuel assemblies loaded in the core, a spectrum shift rod that is provided in the fuel assemblies and whose internal water level changes according to the cooling water flow rate of the core, and is provided in the core so as to be movable in the axial direction. A core performance calculation device for a boiling water reactor that calculates the core characteristics including the power distribution in the reactor by nuclear thermal hydraulic coupling calculation, based on the neutron flux distribution in the actual core The means for calculating the power distribution in the axial direction of the core, the means for calculating the power distribution in the axial direction of the core by the nuclear thermal hydraulic coupling calculation, and the power distribution in the axial direction of both cores are compared, The comparison means for identifying the fuel assembly in which the difference between the water level and the water level set at the time of calculating the core performance has occurred, and the speci fi cation that has been set to match the water level in the actual spectrum shift rod. Characterized in that it comprises means for recalculating corrects the level of torque in the shift rod. Furthermore, the core performance calculation apparatus in the boiling water reactor has a display device for displaying the water level of the spectrum shift rod obtained by recalculation.

  In addition, it is a nuclear reactor plant having the above core performance calculation device in a boiling water reactor.

  According to the present invention, the calculation accuracy of the core characteristics can be further improved in a boiling water reactor loaded with a fuel assembly having SSR.

  Embodiments of the present invention will be described below with reference to the drawings.

(Outline of BWR plant)
A core performance calculation apparatus according to a first embodiment which is a preferred embodiment of the present invention applied to a boiling water reactor (BWR) plant will be described below with reference to FIGS.

  As shown in FIG. 2, in the BWR plant control system, the core performance calculation device 30 forms a network with a data processing computer 31, a storage device 33, etc. via a signal transmission line 36, and information is transmitted between them. We are exchanging. A process input / output device 32 for inputting / outputting data and inputting a control program is connected to the data processing computer 31. The signal transmission line 36 is connected to a display device (CRT) 38 and a display instrument 39 provided in a central control panel 37 in the central operation control room of the BWR plant, and is further connected to a printer 35 and an operator console 34. In FIG. 2, the data processing computer 31 calculates the core flow rate based on the pressure measurement value, and calculates the reactor output based on the neutron flux measurement value. The calculated core flow rate and reactor power are stored in the storage device 33. The core performance calculation device 30 calculates the nuclear constant using the operating state data such as the core flow rate stored in the storage device 33, and executes the core performance calculation. The calculation results obtained by the core performance calculation device 30 and the information calculated by the data processing computer 31 are output to the display device 38, the display instrument 39 and the printer 35, and are displayed and printed.

  Next, a schematic configuration of the BWR plant will be described. As shown in FIG. 3, a nuclear reactor 1 of a BWR plant includes a reactor pressure vessel (hereinafter referred to as RPV) 2 having a core 3 loaded with a fuel assembly. An internal pump 22 is installed at the bottom of the RPV 2. The impeller 23 of the internal pump 22 is disposed in a downcomer 25 that is an annular flow path formed between the RPV 2 and the core 3. A differential pressure gauge 24 that measures the differential pressure between the upstream side and the downstream side of the impeller 23 is installed in the RPV 2. A plurality of control rods 20 inserted into the core 3 are disposed in the RPV 2. The control rod 20 is connected to a control rod drive mechanism 21 installed at the bottom of the RPV 2. The control rod drive mechanism 21 performs an operation of pulling out the control rod 20 from the core 3 and an operation of inserting the control rod 20 into the core 3. Further, a plurality of fixed neutron detectors (local output range monitor) 29 are arranged in the reactor core 3.

  During the operation of the nuclear reactor, the cooling water supplied to the core 3 flows into the fuel assembly and ascends the cooling water flow path formed between the fuel rods. While the cooling water ascends the cooling water flow path, the cooling water is heated by heat generated by nuclear fission of the nuclear fuel material in the fuel rod, and a part thereof becomes steam. In this steam, moisture is removed by a steam separator (not shown) and a steam dryer (not shown) disposed above the core 3 in the RPV 2, and a turbine (not shown) is connected from the RPV 2 by the main steam pipe 26. )). The steam discharged from the turbine is condensed into water by a condenser (not shown) and supplied to the RPV 2 through the water supply pipe 27 as water supply. The cooling water separated by the steam separator is mixed with the feed water supplied into the RPV 2, descends in the downcomer 25, is pressurized by the internal pump 22, and is supplied to the core 3 again.

  Pressure measurement values upstream and downstream of the impeller 23 measured by the differential pressure gauge 24, neutron flux measurement values measured by the mobile neutron detector (TIP), and control rod output from the control rod drive mechanism 21 The position information is input to the data processing computer 31 via the process input / output device 32.

  In FIG. 4, the fuel assembly 4 includes a plurality of fuel rods 5, an upper tie plate 6, a lower tie plate 7, a plurality of fuel spacers 8, a plurality of SSRs 9 and a channel box 19.

  Each fuel rod 5 filled with nuclear fuel material has a lower end portion held by the lower tie plate 7 and an upper end portion held by the upper tie plate 6. The two SSRs 9 are similarly held at the lower tie plate 7 at the lower end and at the upper tie plate 6 at the upper end. The plurality of fuel spacers 8 are arranged at predetermined intervals in the axial direction of the fuel assembly 4, and the fuel rods are formed so as to form cooling water passages between the fuel rods 5 and between the fuel rods 5 and the SSR 9. 5 and SSR9. The bundle of fuel rods 5 bundled by the fuel spacer 8 is surrounded by a channel box 19 attached to the upper tie plate 6.

In the cross-sectional view of the fuel assembly 4 shown in FIG. 5, the plurality of fuel rods 5 are arranged in 9 rows and 9 columns in a square lattice pattern. Reference numeral 10 denotes a riser pipe of the SSR 9, and reference numeral 14 denotes a coolant rise passage. Reference numeral 11 denotes a downcomer, and reference numeral 15 denotes a coolant descending passage.
(SSR operation)
In the schematic diagram of FIG. 6, SSR 9 is an inverted U-shaped tube, and has an ascending tube 10 and a descending tube 11. A coolant ascending passage 14 is formed in the ascending pipe 10, and a coolant descending passage 15 is formed in the descending pipe 11. The coolant rising passage 14 and the coolant lowering passage 15 are communicated with each other at the upper end portion of the SSR 9. The inlet opening 12 formed at the lower end of the ascending pipe 10 communicates with a region in the lower tie plate 7 below the fuel holding portion 28. The lower end of the downcomer pipe 11 is located above the fuel holding portion 28 and below the upper end of the SSR 9, and is preferably located near the upper surface of the fuel holding portion 28. An outlet opening 13 is formed at the lower end of the downcomer pipe 11. The two SSRs 9 are disposed adjacent to the center of the fuel assembly 4 as shown in FIG. The cross-sectional area of the rising pipe 10 of each SSR 9 occupies a region where four fuel rods 5 can be arranged. The cross-sectional area of the downcomer 11 is smaller than the cross-sectional area of the ascending pipe 10. The lower part of the fuel rod 5 is held on the fuel holding part 28.

  Next, the cooling water in the SSR 9 will be described in detail. A part of the cooling water flowing into the lower tie plate 7 of the fuel assembly flows into the coolant rising passage 14 through the inlet opening 12. The cooling water in the coolant ascending passage 14 is heated by irradiation with neutrons and γ rays from the fuel rods held around the SSR 9 and partly evaporates. The steam rises in the coolant ascending passage 14, is discharged from the outlet opening 13 through the coolant descending passage 15, and rises in the coolant flow path between the fuel rods.

  When the core flow rate is less than the maximum core flow rate and the reactor power is in the vicinity of the rated output, the supply flow rate of cooling water into the coolant rising passage 14 and the amount of steam generated in the coolant rising passage 14 are balanced. As a result, the liquid level 16 is formed in the coolant rising passage 14. Z is the water level in the SSR indicating the water level of the coolant from the lower part of the coolant ascending passage to the liquid level 16. Below the liquid level 16 is a water region 17 and above the liquid level 16 is a vapor region 18. Next, when the core flow rate is increased by the internal pump 22, the liquid level 16 in the coolant ascending passage 14 rises, and when the core flow rate reaches the maximum core flow rate, the coolant ascending passage 14 and the coolant descending passage 15. Inside is the water region 17.

When the liquid level 16 is formed in the coolant rising passage 14 in the _SSR 9, the pressure loss ΔP _{SSR in the SSR 9} can be calculated by the equation (1).

ΔP _SSR = ΔP _IN + ΔP _Z + ΔP _OUT (1)
Here, ΔP _IN is a local pressure loss at the inlet opening 12 of the SSR 9, ΔP _Z is a static water head of the cooling water in the SSR 9, and ΔP _OUT is a local pressure loss at the outlet opening 13 of the SSR 9.

When the liquid level 16 is formed in the SSR 9, the flow rates of the cooling water flowing into the SSR 9 and the steam flowing out from the SSR 9 (or steam and cooling water) are small, and ΔP _IN and ΔP _OUT are negligible compared to ΔP _Z. It gets smaller.

On the other hand, the fuel holding portion 28 of the lower tie plate 7 is formed with a large number of openings 7a for supplying cooling water from the lower region to the upper region, and serves as a resistance against the flow of the cooling water. When the nuclear reactor 1 is operated, most of the cooling water flowing into the lower tie plate 7 is introduced into the cooling water passage formed between the fuel rods 5 through the opening 7a. When the lower tie plate pressure loss at the opening 7a is ΔP _LTP , the pressure loss ΔP _SSR and ΔP _LTP of the _{SSR 9} are balanced during the normal operation of the reactor 1 except during the transient change. Further, as described above, ΔP _IN and ΔP _OUT are smaller than ΔP _Z and can be omitted, and equation (2) is established approximately.

ΔP _LTP = ΔP _Z (2)
Here, when the flow rate of the cooling water supplied to the lower tie plate 7, that is, the core flow rate is changed, ΔP _LTP that is the differential pressure between the upstream and downstream regions from the fuel holding portion 28 described above changes. Since ΔP _LTP is proportional to the square of the core flow rate, for example, if the core flow rate is changed from 80% to 120% of the rated flow rate, the lower tie _plate pressure loss ΔP _LTP is about 2.3 times that at 80% flow rate. become. Hydrostatic head ΔP in SSR9 _Z also becomes about 2.3 times in the same manner, the liquid level 16 in the SSR9 greatly increases.

  When the liquid level 16 in the SSR 9 rises, a part of the steam region 18 in the coolant rising passage 14 becomes the water region 17. In the area that newly becomes the water region 17, the moderator density of the cooling water having the neutron moderating function increases, so that thermal neutrons increase and the core reactivity increases. On the contrary, when the liquid level 16 in the SSR 9 is lowered, a part of the water region 17 in the coolant rising passage 14 becomes a steam region 18. In the region that has newly become the steam region 18, thermal neutrons decrease due to a decrease in moderator density, and the core reactivity decreases.

As described above, compared with control rod reactivity control, reactivity control by SSR has an advantage in terms of fuel economy because less neutrons are wastedly absorbed. On the other hand, in order to manage the operation of the reactor using SSR, it is necessary to perform core management in consideration of the moderator density in the SSR. Alternatively, the core ratio can be managed in consideration of the void ratio in the SSR by using the void ratio, which is an index that is 1 for the vapor single phase and 0 for the water single phase, with the volume ratio of the gas phase to the flow path cross-sectional area. Need to be implemented.
(Core performance calculation method)
Details of the core performance calculation method executed by the core performance calculation apparatus 30 will be described below using the flowchart of FIG. The processing of steps 51 to 63 shown in FIG.

  In the first embodiment, first, the entire core to be subjected to nuclear constant calculation is divided into a plurality of nodes of about 15 cm cubic, for example, in the axial direction along the boundary of the fuel assembly with the effective fuel length as a target. It is assumed that each node is homogeneous, and the nuclear constant is calculated. Next, the neutron distribution in the core 3 and the effective neutron multiplication factor are obtained by neutron diffusion calculation. Based on this, core characteristics such as thermal power distribution and reactivity in the core 3 are further calculated.

In the boiling water reactor, the moderator density distribution (void ratio distribution) changes depending on the power distribution of the core 3, so the core performance calculation device 30 repeatedly performs the neutron diffusion calculation process and the thermal hydraulic calculation process. Power distribution and void ratio distribution that are consistent with each other are obtained by nuclear thermal hydraulic coupling calculation.
(Initial setting: Steps 51 and 52)
In step 51, operating state data is input to the core performance calculation apparatus 30. Next, in step 52, initial setting such as axial power distribution of the core 3 is performed based on the operation state data. The operation state data includes the burnup, the moderator density in the fuel assembly 4 outside the SSR 9, the moderator density in the SSR 9, information on the control rod insertion status, the core flow rate, the core inlet temperature, and the like. The control rod insertion state information (control rod position information) and the core flow rate are information obtained by the data processing computer 31.

  The burnup is an index indicating the accumulated output per unit uranium weight, and is obtained by time integration of the node output calculated by the core performance calculation device 30. The burnup value inherits the value calculated by the core performance calculation device 30 and stored in the storage device 33 at the end of the previous operation cycle, and is updated during the current operation cycle.

The moderator density inside the fuel assembly 4 outside the SSR 9 is calculated using the density of water and steam at the operating pressure after calculating the void fraction of the node inside the fuel assembly 4 outside the SSR 9 in the thermal hydraulic calculation. To do. The average combustion moderator density at the node is the moderator density generated at the node in the fuel assembly 4 from the first loading of the core to the present (the moderator density outside the SSR9 and the moderator density in the SSR9) Time average value). The combustion average moderator density is also calculated by the core performance calculation device 30, and the value stored in the storage device 33 is taken over and updated during the current operation cycle.
(Thermal hydraulic calculation processing: steps 53 to 55)
Next, at step 53, the coolant flow rate distribution calculation for each fuel assembly 4 channel is performed, the void ratio distribution calculation within the fuel assembly 4 is performed at step 54, and the hot water including the water level calculation within the SSR is performed at step 55. Force calculation processing is executed.

  In the flow rate distribution calculation of step 53, the total core flow rate obtained by the data processing computer 31 and the output distribution information of the fuel assemblies 4 in the core 3 are used so that the pressure loss of each fuel assembly 4 becomes equal. The flow rate of the cooling water to be supplied is obtained for each fuel assembly 4. Calculation of the pressure loss of the fuel assembly 4 is performed as follows. Assuming the distribution of coolant flow in each fuel assembly, the generated steam volume for each node is calculated using the power distribution and core inlet temperature set in step 52, and the steam flow rate (total mass flow rate of water and steam combined) is calculated. Calculate the axial distribution of the ratio of vapor mass flow to The pressure loss of the node can be calculated from the coolant flow rate and the steam flow rate, and the total pressure loss for each fuel assembly can be calculated by integrating the pressure loss of each node in the axial direction. In the initially assumed cooling water flow distribution, the total pressure loss of each fuel assembly takes a different value for each assembly, but the cooling water flow distribution of each fuel assembly is readjusted to match this, and the pressure loss calculation is performed. repeat. When the total pressure loss of all the fuel assemblies finally matches, the flow distribution calculation 53 ends.

In the void ratio distribution calculation in step 54, the void ratio distribution in the fuel assembly 4 is calculated outside the SSR 9. The steam flow rate of the node is calculated using the power distribution set in step 52, the temperature at the core inlet of the cooling water (core inlet temperature) and the cooling water flow rate of each fuel assembly calculated in step 53. The relationship between the steam weight ratio and the void ratio, which is the steam volume ratio, is obtained in advance by experiments or the like, and the axial distribution of the void ratio α _c (moderator density ρ _c ) is calculated from the axial distribution of the steam weight ratio.

In the water level calculation in the SSR 9 in step 55, the water level Z in the SSR is calculated using the equation (1). First, the lower tie _plate pressure loss ΔP _LTP is calculated from the flow rate W of the cooling water supplied to the lower tie _plate 7 obtained in step 53.

ΔP _LTP = K _LTP · W ² / (ρ _f A ² ) (6)
Here, K _LTP · is the lower _{tie plate} pressure loss coefficient, and A is the flow area of the lower tie plate portion. By substituting equation (6) into equation (1), the following equation (7) for obtaining the water level Z in the SSR from the flow rate W of the cooling water supplied to the lower tie plate 7 is obtained. The lower _tie plate pressure loss coefficient _KLTP and the channel area A of the lower _tie plate portion are obtained in advance through experiments or the like and stored in the storage device 33.
Z = (K _LTP · W ² ) / ((ρ _f ) ² gA) (7)
Under the condition that the water density ρ _f is constant, the water level Z in the SSR is proportional to the square of the flow rate W of the cooling water supplied to the lower tie plate 7, so the equation (7) is corrected and the following equation (8 ) Can also be used.
Z = β (ρ _f ) · W ² (8)
Here, β (ρ _f ) is a correlation coefficient between the water level Z in the SSR at the water density ρ _f and the flow rate W of the cooling water supplied to the lower tie plate 7. β (ρ _f ) is obtained in advance by experiments or the like and stored in the storage device 33 as a table for the water density ρ _f .

After obtaining the water level Z in the SSR by the equation (7) or the equation (8), the water level in the SSR and the positional relationship between the nodes in the axial direction are obtained, and the void ratio α _s in the SSR of each node is obtained. First, a node whose upper end is higher than the water level Z in the SSR and whose lower end is lower than the water level Z in the SSR (hereinafter referred to as a node having a water level) is determined. The upper end and lower end positions of each node are obtained during node division and stored in the storage device 33. At nodes below the water level node, the SSR is all water, so the void rate α _s is 0%. In addition, at the node above the node having the water level, the SSR is entirely water vapor, so the void rate α _s is set to 100%. The void ratio α _s of the node having the SSR water level is obtained by averaging the following equation (9) using the water level and the upper end position U and the lower end position L of the node.

α _s = (U−Z) / (UL) × 100 (9)
In steps 54 and 55, the void ratio α _c in the fuel assembly 4 and the void ratio α _s in the SSR 9 are obtained outside the SSR 9 at each node, respectively, and then the following expressions (10) and (11) are used. Thus, the moderator density ρ _c in the fuel assembly 4 and the moderator density ρ _s in the SSR 9 are calculated outside the SSR 9.

ρ _c = α _c ρ _g + (1-α _c ) ρ _f (10)
ρ _s = α _s ρ _g + (1−α _s ) ρ _f (11)
Here [rho _g is vapor density, the [rho _f is water density.
(Nuclear constant calculation: Step 56)
Next, in order to combine the nuclear calculation in the neutron diffusion calculation process and the above-described thermal hydraulic calculation, the nuclear constant at each node is obtained. That is, in the nuclear constant calculation process of step 56, the node is determined based on the moderator density ρ _c in the fuel assembly 4 and the moderator density ρ _s in the SSR 9 and the control rod insertion state outside the SSR 9 of each node. Find the average nuclear constant.

Examples of the nodal average nuclear constant used in this embodiment include a diffusion coefficient, a macro absorption cross section, a macro scattering cross section, and a macro fission cross section. These nuclear constants are calculated in several representative states by assembly calculations for a single fuel assembly using burnup, moderator density ρ _c and ρ _s , control rod insertion state information, core flow rate, etc. The value is calculated in advance. As the moderator density ρ _c and ρ _s , the values calculated by the equations (4) and (5) are used.

The calculated nuclear constants are associated with the burnup, moderator density ρ _c and ρ _s , control rod insertion state, core flow rate, etc. (hereinafter referred to as second state quantities) used for the calculation, respectively. It is stored in the storage device 33 as information.

In step 56, the burnup, moderator density ρ _c and ρ _s calculated for one fuel assembly 4 that is the object of neutron diffusion calculation, control rod insertion state, core flow rate, etc. (hereinafter referred to as the first state quantity). The average nuclear constant at each node is retrieved from the storage device 33. If there is no second state quantity that matches each value of the first state quantity in the storage device 33, the storage apparatus stores the second state quantity closest to each value of the first state quantity and each corresponding nuclear constant. Each average nuclear constant for the node is obtained by interpolating (or extrapolating) the retrieved second state quantity and each nuclear constant using each value of the first state quantity. Similarly, the average nuclear constants are obtained for the other fuel assemblies.
(Neutron diffusion calculation: Steps 57 and 58)
Next, using the nuclear constant obtained in step 56, neutron diffusion calculation is performed in step 57, and core characteristics (information on the core characteristics) such as the axial power distribution and reactivity of the core are calculated. In step 58, the axial output distribution obtained in the previous step is compared with the axial output distribution initially set in step 52, and a convergence determination is made as to whether or not the difference is within an allowable range.

If the difference in convergence determination is larger than the allowable range, the determination in step 58 is “No”, the output distribution is adjusted, and the processes in steps 53 to 58 are executed again. That is, the thermal hydraulic calculation process (steps 53, 54 and 55), the nuclear constant calculation process (step 56) and the neutron diffusion calculation process (step 57) are repeatedly executed until the determination in step 58 becomes “Yes”. . In the void ratio distribution calculation (step 54) in this iterative calculation, the void ratio distribution is calculated using the axial output distribution obtained in step 57. In addition, in the convergence determination at step 58 in the case of repeated calculation, the difference between the axial output distribution obtained by the previous calculation at step 57 and the axial output distribution obtained by the calculation at step 57 is compared. To do. The axial output distribution and the like obtained in step 57 may be calculated using neutron transport calculation instead of neutron diffusion calculation.
(Axial output distribution: steps 59, 60)
When the determination of step 58 is “Yes”, that is, when the above-described difference is within the allowable range, the comparison process of step 59 is executed.

  As shown in FIG. 7, the process of step 59 is performed by the axis of the core 3 obtained from the axial neutron flux distribution of the core 3 measured by the mobile neutron flux detector (TIP) 40 inserted into the core 3. Use directional output distribution. In the BWR, the TIP 40 does not always measure the neutron flux distribution in the core 3, but is inserted into the core 3 based on the insertion command input from the operator console 34 by the operator, and the neutron flux in the axial direction of the core 3. Measure the distribution. The TIP 40 is pulled out of the reactor 1 during a period other than measuring the neutron flux distribution in the core 3. The measured neutron flux values at a number of positions in the axial direction of the core 3 output from the TIP 40 are input to the data processing computer 31 via the process input / output device 32. The data processing computer 31 obtains the power distribution in the axial direction of the core 3 based on the measured neutron flux values. The power distribution in the axial direction of the core 3 (hereinafter referred to as a TIP measurement value) obtained by the data processing computer 31 is input to the performance calculation device 30. The input TIP measurement value is compared with the axial output distribution obtained by the nuclear thermal hydraulic coupling calculation, and the difference is calculated.

  In step 60, it is determined whether the difference between the TIP measurement value calculated in step 59 and the axial output distribution obtained by the nuclear thermal hydraulic coupling calculation is within an allowable range. If the difference is outside the allowable range, the water level correction in the SSR is performed (step 61).

In the core performance calculation, when there is a difference in the power distribution evaluated by the TIP measurement value and the nuclear thermohydraulic coupling calculation, correction is performed to match the TIP measurement value with the nuclear thermohydraulic coupling calculation to improve the evaluation accuracy. improves. Here, since the TIP 40 is arranged so as to be surrounded by four fuel assemblies as shown in FIG. 7A, the TIP measurement value is an average output distribution of the four fuel assemblies surrounding the TIP 40. It becomes. If any of the four fuel assemblies surrounding the TIP 40 has a difference between the water level in the SSR calculated in step 55 and the actual water level in the SSR, this difference in water level is the cause. There is a difference between the TIP measurement and the power distribution evaluated by the nuclear thermal hydraulic combined calculation. Therefore, in order to correct the output distribution with high accuracy, it is necessary to specify the fuel assembly in which the water level difference in the SSR has occurred and correct the water level in the SSR.
(SSR water level correction: Step 61)
The present invention focuses on the point that the water level in the SSR differs depending on the flow rate of the cooling water supplied to the fuel assembly, and the flow rate of the cooling water supplied to the fuel assembly differs depending on the output of the fuel assembly. That is, in the flow distribution of the cooling water to each fuel assembly, if the output of the fuel assembly is high and the steam flow rate in the fuel assembly becomes high, the pressure loss increases, so the cooling to the fuel assembly Water flow is relatively reduced. Accordingly, the water level in the SSR of the fuel assembly having a high output is relatively low. Normally, the fuel is loaded in the core so that fuels with relatively high outputs or fuels with relatively low outputs are not adjacent vertically or horizontally so that only the output of some fuel assemblies does not become extremely high. Set the pattern. Therefore, in a normal core, fuels having different outputs are often adjacent to each other.

  FIG. 7B schematically shows the SSR water levels of the four fuel assemblies A, B, C, and D surrounding the TIP 29 shown in FIG. 7A. The output of each fuel assembly is B> D> A> C, and the water level in the SSR is C> A> D> B. FIG. 7 (c) shows a difference between the fuel assemblies A and C, such as the actual SSR water level indicated by the solid line in FIG. 7 (b) and the water level in the SSR in the nuclear thermal hydraulic coupling calculation indicated by the dotted line. Shows the difference in power distribution between TIP measurement and nuclear thermohydraulic coupling calculation.

  A node that is above the actual water level in the SSR and that is below the calculated water level in the SSR in the nuclear thermohydraulic coupling calculation, that is, the fuel assemblies A and C in FIG. It was found that the output by the nuclear thermal hydraulic coupling calculation is locally higher than the TIP measurement value in the node including the area surrounded by the dotted line. Also, due to the difference in the SSR water levels of fuel assemblies A and C, the difference between the TIP measurement value and the output from the nuclear thermal hydraulic coupling calculation due to the difference between the actual water levels in the SSR It turns out that it can be identified as a difference. Therefore, there is a difference between the actual SSR water level and the SSR water level in the nuclear thermal hydraulic coupling calculation from the axial position where the output difference between the TIP measurement value and the nuclear thermal hydraulic coupling calculation occurs. It can be specified that the aggregates are A and C.

  7 (b) and 7 (c), the fuel assemblies A and C show examples in which the SSR water level in the nuclear thermal hydraulic coupling calculation is higher than the actual SSR water level. If the water level in the SSR in the nuclear thermohydraulic coupling calculation is lower than the actual water level in the SSR, the output from the nuclear thermohydraulic coupling calculation is locally lower than the TIP measurement value. It is possible to identify the fuel assemblies that have a difference in the water level in the SSR in the nuclear thermal hydraulic coupling calculation.

  In step 61, the procedure for correcting the water level in the SSR using the above principle will be described in detail with reference to the flowchart of FIG. First, in step 61a, a node Y where the difference between the TIP measurement value and the output from the nuclear thermal hydraulic coupling calculation is locally large is identified. Here, a node having a locally large difference in output is, for example, one or two nodes in an axial region where the output differs by 1% or more with reference to the axial height of one node in the nuclear thermal hydraulic coupling calculation. Thus, the output difference is less than 1% at the node above and below the node.

  Next, in step 61b, among the four fuel assemblies surrounding the TIP 29, a fuel assembly in which the water level Z in SSR obtained in step 55 is close to the node Y identified in step 61a is selected. Specifically, when the output of the node Y by the nuclear thermal hydraulic coupling calculation is larger than the output of the node Y in the TIP measurement value, for example, the water level Z in the SSR is above the upper end of the node Y, The fuel assembly closest to the upper end of the node Y is selected. Conversely, when the output of node Y by the nuclear thermal hydraulic coupling calculation is smaller than the output of node Y in the TIP measurement value, for example, the water level Z in the SSR is lower than the lower end of node Y, and node Y Select the fuel assembly closest to the bottom edge of

Next, in step 61c, the SSR water level Z of the selected fuel assembly is corrected. Specifically, when the output of the node Y by the nuclear thermal hydraulic coupling calculation is larger than the output of the node Y in the TIP measurement value, the equation ( The lower _{tie plate} pressure loss coefficient K _{LTP in} 7) or the correlation coefficient β (ρ _f ) in equation (8) is decreased by ΔK or Δβ. Conversely, when the output of node Y by the nuclear thermal hydraulic coupling calculation is smaller than the output of node Y in the TIP measurement value, equation (7) is set so that the water level Z in the SSR of the selected fuel assembly becomes higher. The lower _{tie plate} pressure loss coefficient K _LTP or the correlation coefficient β (ρ _f ) of the equation (8) is increased by ΔK or Δβ. For ΔK or Δβ, for example, a value is obtained in advance so that the water level in the SSR for the same cooling water flow rate changes by ½ of the height of the node, and is stored in the storage device 33. The corrected lower _tie plate pressure loss coefficient K _LTP ′ or correlation coefficient β (ρ _f ) ′ is stored in the storage device 33.

  In step 61d, the difference between the TIP measurement value and the output from the nuclear thermohydraulic coupling calculation is locally increased, that is, due to the difference between the actual SSR water level and the SSR water level in the nuclear thermohydraulic coupling calculation. Therefore, the nodes that can be considered to have a difference in output between the TIP measurement value and the nuclear thermohydraulic coupling calculation are identified for all fuel assemblies for which the TIP measurement value is obtained, and the fuel that causes the output difference Determine whether the water level in the SSR of the aggregate has been corrected.

When “Yes” is determined in the step 61d, the process returns to the step 55. In step 55, the water level in the SSR is obtained using the lower _{tie plate} pressure loss coefficient K _LTP 'or the correlation coefficient β (ρ _f )' corrected in step 61. Then, as before the correction, the nuclear thermal hydraulic coupling calculation is repeated until the output distribution converges. If “Yes” is determined in the convergence determination in step 58, the TIP measurement value is compared with the output distribution by the nuclear thermal hydraulic coupling calculation to obtain a difference (step 59), and it is determined whether the difference is within the allowable range. (Step 60).

When the difference is outside the allowable range, the water level correction in the SSR is performed again (step 61). However, the correction amount ΔK ′ or Δβ ′ of the lower tie plate pressure loss coefficient K _LTP or the correlation coefficient β (ρ _f ) is made smaller than the previous correction amount ΔK or Δβ, for example, the previous correction amount ΔK or Δβ, respectively. Is multiplied by 0.8. After correcting the water level in the SSR, the nuclear thermal hydraulic coupling calculation is performed again. This iterative calculation is repeatedly executed until the difference in output distribution between the TIP measurement value and the nuclear thermal hydraulic coupling calculation is within the allowable range, and “Yes” in step 60.
(Core safety evaluation: steps 62 and 63)
If “Yes” in step 60, the safety evaluation of the core is performed in step 62. In step 62, indices indicating the safety of the core 3 (for example, the minimum limit power ratio and the maximum linear power density) are calculated for the calculated core characteristics. The calculation of the minimum limit output ratio and the like in step 62 is also performed when the operator inputs a calculation command to the core performance calculation apparatus 1 from the operator console 34. After the processing of step 62 is completed, information on the core characteristics such as the power distribution and reactivity of the core in step 63 and each index indicating the calculated safety of the core are displayed from the core performance calculation device 30 to the display device 38 and the display. It is output to the instrument 39. Information on each core characteristic is displayed on the display device 38 or the like. Similarly, the water level in the SSR 9 is also output to the display device 38 and the like. In addition, information on each core characteristic, the water level in the SSR 9, and each index indicating the safety of the core are stored in the storage device 33. The operator makes a nuclear reactor operation plan such as setting the insertion / extraction position of the control rod 20 into the core 3 in consideration of the displayed power distribution and core characteristics such as reactivity.

  By the way, in step 61a of step 61 described above, when identifying a node having a large difference in output distribution by the TIP measurement value and the nuclear thermohydraulic coupling calculation, the fact that the difference is locally large indicates that the actual water level in the SSR and the nucleus The condition for judging that there was a difference with the water level in the SSR in the thermo-hydraulic coupling calculation. However, for example, in the unlikely event that water vapor flows from the connecting part of the SSR riser and downcomer due to an unexpected factor, all the water in the SSR becomes water, and above the water level in the SSR in the nuclear thermal hydraulic coupling calculation. For all nodes, there will be a difference between the TIP measurement and the output from the nuclear thermal hydraulic coupling calculation.

  Therefore, in step 61a, for example, there is a difference of 1% or more between the TIP measurement value and the output of the nuclear thermal hydraulic coupling calculation for all the nodes above the water level in the SSR in the nuclear thermal hydraulic coupling calculation. In this case, it can be determined that the water level is no longer formed in the SSR of the fuel assembly, and the information can be displayed on the display device 38 or the display instrument 39. As a result, the operation plan of the nuclear reactor such as setting of the insertion / extraction position of the control rod 20 into the core 3 is taken into consideration by the operator considering that the water level is not formed in the SSR of the fuel assembly. The convenience of operation management is improved.

  In Example 1, since the flow rate of the cooling water supplied to the fuel assembly changes according to the change in the output of the fuel assembly and the water level in the SSR changes, the output distribution of the TIP measurement value and the nuclear thermal hydraulic coupling calculation is locally A method for identifying the fuel assemblies that have a difference between the actual water level in the SSR and the water level in the SSR in the thermo-hydraulic coupling calculation from different axial positions is presented.

  In the second embodiment, even when the SSR water level is approximately the same among the fuel assemblies, the fuel assembly in which a difference is generated between the actual SSR water level and the SSR water level in the thermo-hydraulic coupling calculation is specified. The method is shown.

  FIG. 9 (a) shows the burnup dependence of the reactivity change when the inside of the SSR is changed from water vapor to water. The fuel assembly A and fuel assembly B shown in FIG. 9 (a) have different burnups due to the difference in the number of operation cycles in the furnace. For this reason, the reactivity change when the inside of the SSR changes from water vapor to water Is also different. The burnup is A <B, and the reactivity change is also A <B. Fig. 9 (b) shows the measured TIP values for the fuel assemblies A and B shown in Fig. 9 (a) when the water level in the SSR in the nuclear thermal hydraulic coupling calculation is higher than the actual water level. It shows the difference of the output by the nuclear thermal hydraulic coupling calculation. As shown in Fig. 9 (b), the fuel assembly A, which has a smaller reactivity change when the inside of the SSR is changed from water vapor to water, has a smaller difference between the TIP measurement value and the output from the nuclear thermal hydraulic coupling calculation. I understood that.

A procedure for correcting the water level in the SSR based on the above principle will be described. First, the first nuclear thermo-hydraulic coupling calculation is performed according to procedures 51 to 58 shown in FIG. 1 as in the first embodiment. Then, in step 59, the difference between the TIP measurement value and the output by the nuclear thermal hydraulic coupling calculation is obtained, and in step 60, whether the difference is within an allowable range is examined. If “No” is determined in step 60, in this embodiment, step 161 shown in FIG. 10 is executed instead of step 61 in FIG.
(Void rate correction: Step 161)
In FIG. 10, in step 161a, as in step 61a, a node Y where the difference between the TIP measurement value and the output from the nuclear thermal hydraulic coupling calculation is locally large is identified.

  Next, in step 161b, as in step 61b, the fuel assembly whose SSR water level is close to the node Y identified in step 161a is selected. The difference from step 61b is that a plurality of fuel assemblies whose SSR water level Z is close to node Y are selected. Specifically, when the output of the node Y by the nuclear thermal hydraulic coupling calculation is larger than the output of the node Y in the TIP measurement value, for example, the water level Z in the SSR is above the upper end of the node Y, Two fuel assemblies are selected from those closest to the upper end of the node Y. Conversely, when the output of node Y by the nuclear thermal hydraulic coupling calculation is smaller than the output of node Y in the TIP measurement value, for example, the water level Z in the SSR is lower than the lower end of node Y, and node Y Select the two fuel assemblies from the one closest to the lower end of.

In step 161c, for each of the two fuel assemblies (referred to as fuel assembly A and fuel assembly B) selected in step 161b, the lower _{tie plate} pressure loss coefficient K _{LTP in} equation (7) or , The correlation coefficient β (ρ _f ) in the equation (8) is increased or decreased by ΔK or Δβ.

A correction pattern A is obtained by correcting the lower _{tie plate} pressure loss coefficient K _LTP or the correlation coefficient β (ρ _f ) of the fuel assembly A. A correction pattern B is obtained by correcting the lower _{tie plate} pressure loss coefficient K _LTP or the correlation coefficient β (ρ _f ) of the fuel assembly B.

For each of the correction patterns A and B, the corrected lower _tie plate pressure loss coefficient K _LTP ′ or correlation coefficient β (ρ _f ) ′ is stored in the storage device 33.
(Nuclear thermal hydraulic calculation: step 162)
After step 161c, nuclear thermal hydraulic coupling calculations 53 to 58 are performed for correction patterns A and B, respectively (step 162). After the nuclear thermo-hydraulic coupling calculation for each correction pattern has converged, that is, when “Yes” is obtained in step 58, the TIP measurement value and the output distribution by the nuclear thermo-hydraulic coupling calculation for each correction pattern are compared. To do.
(Correction pattern selection: Step 163)
At this time, as shown in FIG. 9 (a), the fuel assemblies A and B have different burnups, so the reactivity changes when the SSR changes from water vapor to water. If the shape of the output distribution by the joint calculation is also different, the pattern corrected to approach the actual water level in the SSR has a smaller difference from the TIP measurement value. Therefore, the correction pattern having the smaller difference from the TIP measurement value is selected (step 163).

The output distribution by the nuclear thermal hydraulic coupling calculation with the correction pattern selected in step 163 is compared with the TIP measurement value (step 59), and it is checked whether the difference is within the allowable range (step 60). If “No” in step 60, the process returns to step 161 again to correct the water level in the SSR, and the nuclear thermal hydraulic coupling calculation is repeated for each pattern. At this time, as in the first embodiment, the correction amount ΔK ′ or Δβ ′ of the lower tie plate pressure loss coefficient K _LTP or the correlation coefficient β (ρ _f ) is made smaller than the previous correction amount ΔK or Δβ. For example, a value obtained by multiplying the previous correction amount ΔK or Δβ by 0.8 is used. If “Yes” is determined in step 60, the core safety evaluation (step 62) is performed, and the core characteristic information and the like are output to the display device 38 and the display instrument 39 in the same manner as in the first embodiment. .

  In the second embodiment, the nuclear thermal hydraulic coupling calculation is performed for each correction pattern, so the calculation time is longer than in the first embodiment. However, the four fuel assemblies surrounding the TIP 29 have the same water level in the SSR. However, there is an advantage that the calculation accuracy of the core characteristics can be further improved because the fuel assembly having a difference between the actual SSR water level and the SSR water level in the nuclear thermal hydraulic coupling calculation can be identified.

In the second embodiment, an example in which two fuel assemblies are selected as the correction pattern is shown. However, the number of fuel assemblies can be increased from two.
(Combination of Examples 1 and 2)
It is also possible to combine Example 2 and Example 1. Specifically, in step 161a, a plurality of nodes where the difference between the TIP measurement value and the output from the nuclear thermal hydraulic coupling calculation is locally increased are identified (for example, nodes Y1 and Y2), and identified in step 161b. A plurality of fuel assemblies whose water levels in the SSR are close to each of these nodes are selected. For example, the fuel assemblies having the SSR water level close to the node Y1 are A and C, and the fuel assemblies having the SSR water level close to the node Y2 are B and D. At this time, the correction pattern in step 161c is:
Correction pattern (1):
Node Y1 corrects fuel assembly A and node Y2 corrects fuel assembly B correction correction pattern (2):
At node Y1, fuel assembly A is corrected, and at node Y2, fuel assembly D is corrected correction pattern (3):
At node Y1, fuel assembly B is corrected, and at node Y2, fuel assembly C is corrected correction pattern (4):
There are four patterns, such as correcting fuel assembly B at node Y1 and correcting fuel assembly D at node Y2. In step 162, nuclear thermal hydraulic coupling calculation is performed for each of these four patterns, and in step 163, a correction pattern having the smallest difference from the TIP measurement value in each output distribution is selected. The objective can be achieved.

  Further, in creating the correction pattern, for example, the fuel assemblies A, B, and C may be selected at the node Y1, and the fuel assembly D alone may be selected as the node Y2.

In Examples 1 and 2, assuming that the amount of water vapor generated in the SSR is small, the void ratio is 0% below the water level in the SSR, and the void ratio is 100% above the water level. However, in reality, the void ratio distribution is formed in the SSR in the axial direction due to the generation of water vapor, so in order to further improve the calculation accuracy of the core characteristics, it is necessary to consider the void ratio distribution in the SSR. desirable.
(Void rate table information)
In consideration of the above, in this embodiment, instead of calculating the water level in the SSR using the equation (7) or the equation (8) in step 55 and calculating the void rate in the SSR at each node from the water level in the SSR. The table information of the void rate α _s in the SSR corresponding to the coolant flow rate of the fuel assembly as shown in FIG. 11 is used.

11A shows table information when the coolant flow rate of the fuel assembly is W ₁ , and FIG. 11B shows table information when the coolant flow rate of the fuel assembly is W ₂ larger than W ₁ . In this case, for example, in each node below the node number indicated by 41a in FIG. 11 (a), the SSR is all water, and each node above the node number indicated by 42a in FIG. 11 (a). In the node, it indicates that the entire SSR is water vapor. In each node between the node numbers 41a and 42a, water and water vapor are mixed in the SSR. The same applies to FIG. 11B. Further, when the coolant flow rate of the fuel assembly is W _{2 which} is higher than W ₁ , the water level in the SSR becomes high, so that 41a <41b and 42a <42b.

Such table information is obtained by three-dimensional detailed thermal hydraulic calculation or experiment with respect to the cooling water flow rate of some typical fuel assemblies (hereinafter referred to as the second cooling water flow rate for convenience). The information is created in advance based on the stored information and stored in the storage device 33. In step 55 in the present embodiment, using the coolant flow rate of the fuel assembly calculated in step 53 (hereinafter referred to as the first coolant flow rate for convenience), the storage device 33 stores the SSR in the SSR. The void rate α _s table information is searched. If the second cooling water flow rate matching the first cooling water flow rate does not exist in the storage device 33, the second cooling water flow rate close to the first cooling water flow rate and the corresponding SSR void ratio α _s table information are obtained. By searching from the storage device 33 and using the first cooling water flow rate, the second cooling water flow rate and the SSR void rate α _s table information are interpolated (or extrapolated), thereby the SSR for each node. The void ratio α _s is obtained.

By using these table information, the void ratio α _s in the SSR 9 at an arbitrary axial position can be obtained based on the coolant flow rate of the fuel assembly.

By the way, in the void rate α _s table information shown in FIG. 11A, for example, the void rate α _s is approximated so as to change linearly from node numbers 41a to 42a. However, the change may be approximated so as to change in a quadratic function, for example, or may be approximated so that the gradient changes between the node numbers 41a and 42a. Also, necessarily, the void fraction alpha _s is 1.0 when the cooling water flow rate of the fuel assembly inlet is small, may not be the void fraction alpha _s is 0 when the cooling water flow rate is very large.

As the table information, the moderator density ρ _s in the SSR 9 calculated by the equation (5) may be directly used instead of the void ratio α _s in the SSR 9. Specifically, the table information shown in FIG. 11 is replaced with the relationship between the node number and the moderator density ρ _s in the SSR 9. As a method of giving information on the amount of cooling water existing in the SSR 9, the error when the reactor pressure changes is smaller when the moderator density ρ _s in the SSR 9 is used than when the void rate α _s in the SSR 9 is used. .
(Correction of void ratio table information)
A method for correcting the water level in the SSR in the third embodiment will be described below. The procedure for identifying the node Y where the difference between the output from the TIP measurement value and the nuclear thermal hydraulic coupling calculation is locally large, and the procedure for selecting the fuel assembly close to the node Y where the water level in the SSR is identified are The same as in the second embodiment. In Examples 1 and 2, the water level in the SSR was corrected by correcting the lower tie plate pressure loss coefficient K _LTP or the correlation coefficient β (ρ _f ), but in Example 3, the void ratio α _s in the SSR 9 was corrected. The table information is corrected.

Specifically, when the output of the node Y by the nuclear thermal hydraulic coupling calculation is larger than the output of the node Y in the TIP measurement value, for example, the table information of the void ratio α _s in the SSR 9 is shown in FIG. 41a and 41b in FIG. 11B are corrected so as to be shifted by ΔZ _{L to} the smaller node number. On the contrary, when the output of the node Y by the nuclear thermal hydraulic coupling calculation is smaller than the output of the node Y in the TIP measurement value, for example, the table information of the void ratio α _s in the SSR 9 is shown in FIG. 41a and 41b in FIG. 11B are corrected so as to be shifted by ΔZ _L toward the larger node number.

When the correction amount ΔZ _L of the table information is, for example, a half of the axial height of the node at a certain cooling water flow rate, the hydrostatic head in the SSR is proportional to the square of the cooling water flow rate at other cooling water flow rates. In order to maintain the changing relationship, the cooling water flow rate is obtained in advance by calculation or the like and stored in the storage device 33. In addition, when 41a in FIG. 11 (a) and 41b in FIG. 11 (b) are corrected by ΔZ _L , correction amounts ΔZ _U of 42a in FIG. 11 (a) and 42b in FIG. 11 (b) are also calculated in advance. And is stored in the storage device 33 in one-to-one correspondence with the correction amount ΔZ _{L for} each cooling water flow rate.

After correcting the water level in the SSR, either the nuclear thermal hydraulic coupling calculation is performed in the same manner as in Example 1, or the nuclear thermal hydraulic coupling calculation is performed for each correction pattern in the same manner as in Example 2, and the TIP measurement value and The correction pattern with the smallest output difference is selected. Then, it is determined whether the difference between the TIP measurement value and the output from the nuclear thermohydraulic coupling calculation is within the allowable range. At this time, the correction amount ΔZ _L is made smaller in absolute value than the previous time.

  The above is repeated until the difference between the TIP measurement value and the output from the nuclear thermal hydraulic coupling calculation is within the allowable range. When the output is within the allowable range, the safety evaluation of the core in step 62 is performed. In the same manner as in Example 2, information on the core characteristics is output to the display device 38 and the display instrument 39.

  As described above, even when table information is used to calculate the void ratio in the SSR in consideration of the void ratio distribution in the SSR, there is a difference between the actual water level in the SSR and the water level in the SSR in the nuclear thermal hydraulic coupling calculation. The fuel assembly can be identified and the water level in the SSR can be corrected.

It is a flowchart which shows the process sequence of the core performance calculation performed with the core performance calculation apparatus of Example 1 applied to the BWR plant which is one preferable Example of this invention. It is a block diagram of the operation control system incorporating the core performance calculation apparatus of Example 1 in which the processing procedure of FIG. 1 is performed. It is a block diagram of the BWR plant to which the core performance calculation apparatus of Example 1 is applied. FIG. 4 is a longitudinal sectional view of a fuel assembly having an SSR loaded in the core shown in FIG. 3. It is a cross-sectional view of the fuel assembly shown in FIG. It is a schematic diagram of SSR used for the fuel assembly shown in FIG. It is explanatory drawing which shows the relationship between the water level in SR of a fuel assembly, and axial direction output distribution. It is a flowchart which shows the process sequence of the core performance calculation performed with the core performance calculation apparatus of Example 1 applied to the BWR plant which is an Example of this invention. It is explanatory drawing which shows the relationship between the burnup of a fuel assembly, the burnup change when the inside of the SSR is changed from water vapor to water, and the axial output distribution. It is a flowchart which shows the process sequence of the core performance calculation performed with the core performance calculation apparatus of Example 2 applied to the BWR plant which is another Example of this invention. It is explanatory drawing which shows an example of the table information of the void ratio in SSR with respect to an axial direction node.

Explanation of symbols

  DESCRIPTION OF SYMBOLS 1 ... Reactor, 3 ... Core, 4 ... Fuel assembly, 9 ... Spectral shift rod (SSR), 30 ... Core performance calculation device, 31 ... Data processing computer, 32 ... Process input / output device, 33 ... Mass storage device 38 ... Display device, 40 ... Neutron detector (TIP), Z ... SSR water level

Claims (8)

A plurality of fuel assemblies loaded in the core, a spectrum shift rod that is provided in the fuel assemblies and whose internal water level changes according to the coolant flow rate of the core, and is provided in the core so as to be movable in the axial direction. For boiling water reactors with neutron detectors, the core performance calculation method for boiling water reactors that calculates the core characteristics including the power distribution in the reactor by nuclear thermal hydraulic coupling calculation,
Comparing the core axial power distribution calculated based on the neutron flux distribution in the actual core and the core axial power distribution obtained by the nuclear thermal hydraulic coupling calculation, the actual water level in the spectrum shift rod and Identify fuel assemblies that have a difference in the water level set during the core performance calculation, and recalculate and correct the water level in the already set spectral shift rod to match the actual water level in the spectral shift rod. A core performance calculation method in a boiling water reactor characterized by In the boiling water reactor core performance calculation method according to claim 1,
The nuclear thermal hydraulic coupling calculation distributes the cooling water flow rate to each channel of the fuel assembly, and calculates the nuclear constant and neutron distribution of the core based on the void fraction distribution calculation in the fuel assembly and the water level calculation in the spectrum shift rod. And calculating the core performance including the power distribution in the reactor, and a core performance calculation method in a boiling water reactor. In the method for calculating the core performance of a boiling water reactor according to claim 1 or 2,
A fuel assembly having a difference between the water level in the actual spectral shift rod caused by the difference in the cooling water flow rate and the water level in the spectral shift rod set at the time of calculating the core performance is identified, and the fuel assembly in the actual spectral shift rod is identified. A core performance calculation method in a boiling water reactor, wherein the water level in a spectral shift rod that has already been set is recalculated and corrected to match the water level. In the method for calculating the core performance of a boiling water reactor according to claim 1 or 2,
Based on the difference in reactivity to the water level change in the spectral shift rod due to the difference in the nuclear characteristics of the fuel assembly, the difference between the actual water level in the spectral shift rod and the water level in the spectral shift rod set when calculating the core performance The boiling water reactor is characterized by re-calculating and correcting the water level in the spectral shift rod that has already been set so that the fuel assembly in which the gas is generated is identified and matched with the water level in the actual spectral shift rod Core performance calculation method in Japan. In the core performance calculation method of the boiling water reactor according to any one of claims 1 to 4,
When recalculating the water level in the spectrum shift rod, the boiling water type is characterized in that the table information regarding the void fraction in the spectrum shift rod corresponding to the cooling water flow rate of the fuel assembly calculated and stored in advance is used. Core performance calculation method for nuclear reactors. A plurality of fuel assemblies loaded in the core, a spectrum shift rod that is provided in the fuel assemblies and whose internal water level changes according to the coolant flow rate of the core, and is provided in the core so as to be movable in the axial direction. In the boiling water reactor core performance calculation device, which has a neutron detector and calculates core characteristics including power distribution in the reactor by nuclear thermal hydraulic coupling calculation,
The means for calculating the power distribution in the axial direction of the core based on the actual neutron flux distribution in the core, the means for calculating the power distribution in the axial direction of the core by the nuclear thermal hydraulic coupling calculation, and the power distribution in the axial direction of both cores are compared. The comparison means for identifying a fuel assembly having a difference between the actual water level in the spectrum shift rod and the water level set at the time of calculating the core performance has already been adjusted to match the water level in the actual spectrum shift rod. An apparatus for calculating core performance in a boiling water reactor, comprising means for recalculating and correcting a water level in a set spectrum shift rod. In the core performance calculation apparatus for a boiling water reactor according to claim 6,
An apparatus for calculating core performance in a boiling water reactor, comprising a display device for displaying a water level of a spectrum shift rod obtained by recalculation.   A nuclear reactor plant having a core performance calculation apparatus for a boiling water reactor according to claim 6 or 7.

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