JP6099876B2 - Core analysis program and analysis device - Google Patents

Description

  The present invention relates to a core analysis program and an analysis apparatus for evaluating nuclear characteristics around a core including a core.

  2. Description of the Related Art Conventionally, a core design support apparatus including a fuel assembly calculation unit, a homogeneous cell calculation unit, and a core calculation unit is known as an apparatus for evaluating nuclear characteristics in the core (see, for example, Patent Document 1). ). In this core design support device, the fuel assembly calculation unit calculates the nuclear constant of the fuel assembly, the homogeneous cell calculation unit calculates the nuclear constant for each homogeneous cell based on the nuclear constant, and the core calculation unit calculates the nuclear constant for each homogeneous cell. The physical quantity indicating the physical state of the core is calculated on the basis of the nuclear constant.

JP 2007-147377 A

  However, in the conventional core design support apparatus, since the region for evaluating the nuclear characteristics is in the core, it is difficult to evaluate the nuclear characteristics outside the core. Further, in the conventional core design support apparatus, the fuel whose nuclear characteristics are evaluated is a fuel that is not damaged, that is, a healthy fuel, and thus it is difficult to evaluate the nuclear characteristics of the fuel damaged due to melting or the like.

  Therefore, the present invention provides a core analysis program and an analysis apparatus that can evaluate the nuclear characteristics around the core including the core and can evaluate the nuclear characteristics of the fuel in various physical states. Let it be an issue.

  The core analysis program of the present invention is a core analysis program that can be executed on hardware for evaluating the nuclear characteristics of fuel. In the core analysis program, the area where fuel exists outside the core and the core is set as the analysis area, and the analysis area is divided into elements. In order to calculate a nuclear constant, a parameter including an analysis region modeling step simulated by a three-dimensional model composed of a plurality of blocks and a physical state of healthy fuel that is not damaged and damaged fuel that is damaged fuel Based on the input parameters, the nuclear constant calculation step for calculating the nuclear constant for each block and the nuclear constant calculated in the nuclear constant calculation step are set according to the fuel composition conditions in each block. Core calculation in the analysis area based on the nuclear constant set in each block Characterized by comprising the steps, a.

  According to this configuration, the region where the fuel exists inside the core (inside the reactor) and outside the core (outside the reactor) can be set as the analysis region. For this reason, not only the nuclear characteristics of the fuel in the furnace but also the nuclear characteristics of the fuel outside the furnace can be evaluated. Further, since the nuclear constant can be calculated based on the input parameters including the physical state of the healthy fuel and the damaged fuel, not only the nuclear characteristics of the healthy fuel but also the nuclear characteristics of the damaged fuel can be evaluated. This makes it possible to evaluate the nuclear characteristics of healthy fuel inside the furnace, damaged fuel inside the furnace, healthy fuel outside the furnace, and damaged fuel outside the furnace, and various physical states over a wide range. The nuclear characteristics of the resulting fuel can be evaluated.

  In this case, it is preferable that in the nuclear constant calculation step, the homogeneous nuclear constant homogenized in each block is calculated based on the calculation parameter, and in the core calculation step, the homogeneous nuclear constant is set in each block.

  According to this configuration, the calculation system in which the core calculation is performed with the homogeneous nuclear constant set in each block is simpler than the calculation system in which the core calculation is performed with the nuclear constant set in each block in a non-homogeneous state. Therefore, the calculation time can be shortened, and the calculation for evaluating the nuclear characteristics can be performed at high speed.

  The analysis apparatus of the present invention is characterized in that the above core analysis program can be executed.

  According to this configuration, the nuclear characteristics of the fuel in various physical states can be evaluated over a wide range both inside and outside the furnace.

  According to the core analysis program and the analysis apparatus of the present invention, the nuclear characteristics of fuels in various physical states can be evaluated over a wide range both inside and outside the furnace.

FIG. 1 is a structural diagram schematically showing a nuclear reactor when an abnormal time is assumed. FIG. 2 is an explanatory diagram schematically showing a nuclear reactor that simulates the analysis region. FIG. 3 is a schematic configuration diagram schematically showing the analysis apparatus. FIG. 4 is a flowchart relating to the control operation when the core analysis program is executed. FIG. 5 is a flowchart relating to the control operation when the recritical behavior analysis program is executed.

  Hereinafter, a core analysis program and an analysis apparatus according to the present invention will be described with reference to the accompanying drawings. The present invention is not limited to the following examples. In addition, constituent elements in the following embodiments include those that can be easily replaced by those skilled in the art or those that are substantially the same.

  The core analysis program according to the present embodiment can be executed on the analysis apparatus, and the region where fuel is present inside and outside the core is set as the analysis region, and the neutron flux in the analysis region is calculated to calculate the analysis region. Predict and evaluate the distribution and behavior of neutrons. And based on the analysis result obtained by this core analysis program, the recritical behavior in the reactor at the time of abnormality is evaluated. Note that the evaluation of recritical behavior is to evaluate the behavior of neutrons in transient events such as changes in the fuel system due to fuel withdrawal, inflow of moderator, loss of neutron absorber, and the like.

  FIG. 1 is a structural diagram schematically showing a nuclear reactor when an abnormal time is assumed. As shown in FIG. 1, the nuclear reactor 1 is a boiling water nuclear reactor, and includes a pressure vessel 2 having an inner core 5 and a containment vessel 3 for storing the pressure vessel 2. A plurality of fuel assemblies 6 are accommodated inside the pressure vessel 2, and a core 5 is configured by the plurality of fuel assemblies 6. Each fuel assembly 6 has a plurality of fuel rods, and the fuel rods are configured by loading fuel pellets inside the cladding tube.

  In this core analysis program, the nuclear characteristics inside the core 5 and outside the core 5 at the time when the nuclear reactor 1 is abnormal are evaluated assuming that the nuclear reactor 1 is abnormal. Here, as shown in FIG. 1, the assumed abnormal state of the nuclear reactor 1 is a healthy fuel that is an unbroken fuel rod (fuel assembly 6) and a damaged fuel rod (fuel assembly 6). A certain damaged fuel is in a state of being mixed inside the core 5 and outside the core 5. The damaged fuel is, for example, fuel such as fuel in which fuel is melted (molten fuel) and fuel in which the cladding tube is broken (damaged fuel). Specifically, the abnormal state of the nuclear reactor 1 is that some of the fuel rods of the plurality of fuel assemblies 6 of the pressure vessel 2 are stored as healthy fuel inside the pressure vessel 2, while the other A part of the pressure vessel 2 is stored as damaged fuel. In addition, some of the fuel rods of the plurality of fuel assemblies 6 of the pressure vessel 2 come out as healthy fuel outside the pressure vessel 2, while the other part is damaged fuel outside the pressure vessel 2. It is in a state to come out as.

  Next, the core analysis program P1 will be described with reference to FIGS. FIG. 3 is a schematic configuration diagram schematically showing the analysis apparatus, and FIG. 4 is a flowchart regarding a control operation when the core analysis program is executed. The core analysis program P1 is executed by the analysis device 40. The analysis apparatus 40 includes a control unit 41 that can execute various programs and perform calculations, a storage unit 42 that stores various programs and data, an input unit 43 that includes an input device such as a keyboard, and an output device such as a monitor. The output part 44 comprised by these. In the analysis device 40, the control unit 41 performs the core analysis by executing the core analysis program P <b> 1 stored in the storage unit 42 based on the input parameter input from the input unit 43, and outputs the analysis result to the output unit 44. Output to.

  As illustrated in FIG. 4, the control unit 41 includes an input step S100, an analysis region modeling step S101, a nuclear constant calculation step S102, and a core calculation step S103 by executing the core analysis program P1. The step can be executed.

  The input step S100 is a step in which input parameters relating to core analysis are input via the input unit 43 of the analysis device 40. Here, when the fuel is a healthy fuel, the input parameters for performing the core analysis include parameters representing the physical state of the healthy fuel. For example, the fuel position, the fuel composition, the fuel temperature, and the fuel pellet , The outer diameter of the cladding tube, the thickness of the cladding tube, the pitch of the fuel rods, the temperature of the moderator, and the like. In addition, when the fuel is a damaged fuel, the input parameter includes a parameter indicating a physical state of the damaged fuel, for example, a parameter such as a fuel position, a fuel composition, and a fuel shape. Here, as the shape of the damaged fuel, for example, a massive molten fuel having voids (bubbles) formed therein or a molten fuel in a granular form is assumed. Furthermore, the input parameters include parameters representing the physical state of the neutron absorber in addition to the parameters exemplified above, and are parameters such as the composition of the neutron absorber, the shape of the neutron absorber, and the like. The input parameter is not limited to the above example as long as it is a parameter that is an input condition for core analysis.

  FIG. 2 is an explanatory diagram schematically showing a nuclear reactor that simulates the analysis region. The analysis region modeling step S101 is a step of simulating (modeling) the nuclear reactor 1 with a plurality of blocks B. The simulated region is an analysis region E to be evaluated for nuclear characteristics, and the analysis region E is in the range shown in FIGS. That is, the analysis region E is a region where fuel (sound fuel and damaged fuel) exists inside the core 5 and outside the core 5. As shown in FIG. 2, in the analysis area modeling step S101, the analysis area E is divided into elements, and the analysis area E is modeled by dividing into blocks for each divided element. Therefore, in the analysis region modeling step S101, the analysis region E to be modeled is composed of a plurality of blocks B divided for each element.

  The elements include the physical state of the fuel, the presence / absence of a neutron absorber that absorbs neutrons, the presence / absence of the structural material of the reactor 1, and the like in the analysis region modeling step S101, It is divided into a plurality of blocks B.

  In the analysis region modeling step S101, the shape and composition of each block B divided for each element are simulated. For example, in the analysis region modeling step S101, the block B including the fuel element among the plurality of blocks B is modeled according to the physical state of the fuel. That is, the physical state of the fuel includes a damaged state of the fuel, a geometric shape of the fuel, a composition of the fuel, the presence or absence of a structural material mixed in the fuel, and the like.

  The nuclear constant calculation step S102 is a step of calculating the homogeneous nuclear constant of each block B using the nuclear constant calculation code based on the input parameters. Specifically, in the nuclear constant calculation step S102, physical quantities such as a multi-group neutron flux that is a calculation result of the neutron transport calculation are obtained by executing a neutron transport calculation that is a heterogeneous system based on the input parameters in each block B. calculate. Then, a homogenized nuclear constant in each block B, that is, a homogeneous nuclear constant, is calculated using the calculated multi-group neutron flux as a weight.

  Here, the homogeneous nuclear constant is input data input to the core calculation step S103, and examples of the nuclear constant include a diffusion coefficient, an absorption cross section, a removal cross section, and a generation cross section. That is, nuclear constants that are input data for core calculation are generated by performing nuclear constant calculations using the nuclear constant calculation code.

  In executing the neutron transport calculation, a two-dimensional area may be set for each block B to execute the two-dimensional calculation, or a three-dimensional area may be set to execute the three-dimensional calculation. Moreover, as a calculation code of neutron transport calculation, it is preferable to apply a continuous energy Monte Carlo code, for example. In calculating the homogenized nuclear constant from the multigroup neutron flux calculated in the heterogeneous system, it is preferable to calculate by a calculation method based on the equivalence principle.

  The core calculation step S103 is a step of calculating the physical quantity of each block B using the core calculation code. As the core calculation code, for example, a static characteristic core calculation code or a dynamic characteristic core calculation code is applied. The static characteristic core calculation code is used when evaluating the static characteristics, that is, the nuclear characteristics of the core 5 that does not change over time. For example, the static characteristic core calculation code is used when performing safety evaluation of the core 5. The dynamic characteristic core calculation code is used when evaluating the dynamic characteristic, that is, the nuclear characteristic of the core 5 that changes with time. For example, the dynamic characteristic core calculation code is used when monitoring the core 5 or when evaluating the recritical behavior described later. It is done.

  Specifically, in the core calculation step S103, the homogeneous nuclear constant calculated in the nuclear constant calculation step S102 is set corresponding to each block B, and the physical quantity (core core characteristic value in each block B is set based on the set homogeneous nuclear constant. ) Is calculated. The physical quantity includes critical boron concentration, power distribution, reactivity coefficient, delayed neutron ratio, delayed neutron preceding nuclear decay constant, neutron lifetime, and the like.

  When the above-described core analysis program P1 is executed on the analysis device 40, the analysis device 40 calculates the homogeneous nuclear constant in each block B based on the input parameters using the nuclear constant calculation code, and sets the core calculation code as the core calculation code. By using the calculated homogeneous nuclear constant for each block B and performing core calculation, the nuclear characteristics in the analysis region E are evaluated.

  The core analysis program P1 configured as described above is incorporated in, for example, the recritical behavior analysis program P2 for evaluating the behavior related to the recriticality of the fuel. With reference to FIG. 5, the control operation for executing the recritical behavior analysis using the recritical behavior analysis program P2 will be briefly described. FIG. 5 is a flowchart relating to the control operation when the recritical behavior analysis program is executed. Recritical behavior analysis evaluates recriticality behavior of healthy or damaged fuel by tracking changes in neutrons over time from the start time at the beginning of the time change to the end time at the end of the time change It is analysis.

  As shown in FIG. 5, the recritical behavior analysis program P2 uses an improved quasi-positive approximation model that combines the core analysis program P1 and the one-point reactor model. The recritical behavior analysis program P2 includes an input step S111, a nuclear constant calculation step S112, a homogeneous nuclear constant acquisition step S113, an analysis region modeling step S114, a core calculation step S115, a physical quantity acquisition step S116, a shape function An acquisition step S117, a one-point furnace calculation step S118, an amplitude function acquisition step S119, a neutron flux calculation step S120, a thermal / nuclear property evaluation step S121, and an update step S122 are provided.

  The input step S111 is a step in which input parameters related to the recriticality behavior are input via the input unit 43 of the analysis device 40 in the same manner as step S100 described above. Here, the input parameter for analyzing the behavior of the recriticality is a parameter relating to the displacement of the physical state of the fuel, the neutron absorber, etc., which changes in accordance with the scenario where the recriticality may occur. Specifically, these are parameters when the input parameters in step S100 are changed over time, such as changes in the shape of the healthy fuel and damaged fuel, changes in the composition of the healthy fuel and damaged fuel, and the presence or absence of a neutron absorber. is there. The scenario in which recriticality may occur is, for example, a scenario such as a change in fuel system such as fuel removal, inflow of moderator, loss of neutron absorber, and the like.

  In the analysis region modeling step S114, the analysis region E is simulated by dividing the analysis region E for each element and forming a block for each of the plurality of elements, as in step S101.

  In the nuclear constant calculation step S112, the input parameters input in the input step S111 and the analysis region E modeled in the analysis region modeling step S114, using the nuclear constant calculation code, as in step S102 above. Based on this, the homogeneous nuclear constant is calculated. At this time, in the nuclear constant calculation step S112, various homogeneous nuclear constants that change with the passage of time are calculated based on various input parameters that change every predetermined time.

  In the homogeneous nuclear constant acquisition step S113, various homogeneous nuclear constants calculated in the nuclear constant calculation step S112 are acquired and stored in the storage unit 42 of the analyzer 40. Various homogeneous nuclear constants stored in the storage unit 42 of the analysis device 40 are acquired in the core calculation step S115.

  In the core calculation step S115, similarly to the above-described step S103, using the dynamic characteristic core calculation code, the homogeneous nuclear constant stored in the storage unit 42, the analysis region E modeled in the analysis region modeling step S114, and Based on the above, a predetermined physical quantity in each block B is calculated.

  In the physical quantity acquisition step S116, the predetermined physical quantity calculated in the core calculation step S115 is acquired. Here, the predetermined physical quantity is a physical quantity that becomes an input parameter of the single-point furnace calculation step S118. In the physical quantity acquisition step S116, for example, the reactivity, the delayed neutron ratio, the delayed neutron preceding nuclear decay constant, and the neutron lifetime are acquired. Is done.

  In the shape function acquisition step S117, the shape function as the predetermined physical quantity calculated in the core calculation step S115 is acquired.

  In the single point furnace calculation step S118, a predetermined physical quantity is calculated based on the physical quantity acquired in the physical quantity acquisition step S116 using a calculation code to which the single point furnace dynamic characteristic model is applied. The one-point reactor dynamic characteristic model is a model of the reactor 1 that is one-dimensional at an arbitrary point.

  In the amplitude function acquisition step S119, the amplitude function as the predetermined physical quantity calculated in the one-point furnace calculation step S118 is acquired.

  In the neutron flux calculation step S120, the neutron flux is calculated by calculating the product of the shape function acquired in the shape function acquisition step S117 and the amplitude function acquired in the amplitude function acquisition step S119.

  In the thermal / nuclear characteristic evaluation step S121, the nuclear characteristic of the analysis region E is evaluated based on the neutron flux calculated in the neutron flux calculation step S120. That is, in the thermal / nuclear characteristic evaluation step S121, based on the calculated neutron flux, as thermohydrodynamic characteristics, for example, heat generation due to nuclear fission, decay heat, distribution of voids (bubbles generated around the fuel rod) due to radiolysis, temperature, etc. The distribution, output, etc. are calculated, and the calculation result is output to the output unit 44. Further, as the nuclear characteristics, for example, the neutron flux distribution in the analysis region E, the output, the number of fission, and the like are calculated, and the calculation result is output to the output unit 44.

  In the update step S122, the time is updated by a predetermined time in order to evaluate the nuclear characteristics of the analysis region E at every elapse of the predetermined time from the start time to the end time. And after execution of update step S122, it progresses to step S115 again, and the core calculation based on the time after update and the thermal and nuclear characteristic calculated by thermal and nuclear characteristic evaluation step S121 is performed. That is, in the core calculation step S115, using the core calculation code, the homogeneous nuclear constant corresponding to the updated time is acquired from the storage unit 42, and the acquired homogeneous nuclear constant is set in each block B of the analysis region E. Calculate a predetermined physical quantity. In addition, in update step S122, it is repeatedly performed from the start time to the end time.

  In this way, in the recritical behavior analysis program P2, the analysis that changes every predetermined time is performed by performing the core calculation while changing the homogeneous nuclear constant every time the predetermined time elapses from the start time to the end time. The recriticality behavior of region E can be analyzed.

  As described above, according to the configuration of the present embodiment, the region where the fuel exists inside the core 5 and outside the core 5 can be set as the analysis region E. For this reason, not only the nuclear characteristics of the fuel inside the core 5 but also the nuclear characteristics of the fuel outside the core 5 can be evaluated. Further, since the nuclear constant can be calculated based on the input parameters including the physical state of the healthy fuel and the damaged fuel, not only the nuclear characteristics of the healthy fuel but also the nuclear characteristics of the damaged fuel can be evaluated. This makes it possible to evaluate the nuclear characteristics of healthy fuel inside the furnace, damaged fuel inside the furnace, healthy fuel outside the furnace, and damaged fuel outside the furnace, and various physical states over a wide range. The nuclear characteristics of the resulting fuel can be evaluated.

  Further, according to the configuration of the present embodiment, since the core calculation can be performed by setting the homogeneous nuclear constant to each block B, it is simpler than the core calculation using the non-homogeneous nuclear constant. can do. Thereby, the analysis apparatus 40 can shorten the calculation time required for the analysis, and can speed up the calculation for evaluating the nuclear characteristics.

  In this embodiment, the present invention is applied to a boiling water reactor. However, the present invention may be applied to a pressurized water reactor and is not particularly limited.

DESCRIPTION OF SYMBOLS 1 Reactor 2 Pressure vessel 3 Containment vessel 5 Core 6 Fuel assembly 40 Analysis apparatus 41 Control part 42 Memory | storage part 43 Input part 44 Output part E Analysis area B Block P1 Core analysis program P2 Recritical behavior analysis program

Claims (3)

In the core analysis program for evaluating the nuclear characteristics of fuel, which is executed by the analysis device as hardware ,
An analysis region modeling step in which an area where fuel exists inside and outside the core is an analysis area, and the analysis area is simulated by a three-dimensional model including a plurality of blocks divided for each element;
Parameters including the physical state of healthy fuel that is not damaged and damaged fuel that is damaged fuel are used as input parameters for calculating the nuclear constant. A nuclear constant calculation step for performing the calculation;
The nuclear constant calculated in the nuclear constant calculation step is set corresponding to the fuel composition condition in each block, and the core calculation in the analysis region is performed based on the nuclear constant set in each block. Performing the core calculation step to be performed by the analyzer ,
In the analysis region modeling step, the core analysis program is characterized in that the block including the fuel element is modeled according to the physical state of the fuel including the fuel damage state among the plurality of blocks. In the nuclear constant calculation step, a homogenous nuclear constant homogenized in each block is calculated based on the input parameters,
The core analysis program according to claim 1, wherein in the core calculation step, the homogeneous nuclear constant is set in each block. Analysis apparatus and executes the core analysis program according to claim 1 or 2.

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