JP2005338042A - Critical safety design program for equipments for transporting and storing spent fuel - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide critical safety design program for a transportation cask for spent fuel and a storage rack for the spent fuel adopting a rational burn-up credit. <P>SOLUTION: A burn-up of a fuel assembly calculated by a fuel design code based on a collision probability method and a void rate dependent nuclear constant are converted into a transportation cross-section, a portion corresponding to the burn-up of the fuel assembly obtained from a reactor core process computer and an average void rate provided by averaging the void rates in the whole combustion period is extracted out of the transportation cross-section to prepare a transportation cross-section of a spent fuel assembly, and noncritical degrees of the equipments for transporting and storing the spent fuel assembly are calculated by an energy multigroup Monte Carlo method, using arrangement when the transporting and storing the spent fuel assembly, and the transportation cross-section of the spent fuel assembly. <P>COPYRIGHT: (C)2006,JPO&NCIPI

Description

  The present invention relates to criticality safety design of equipment related to transportation and storage of spent fuel, and in particular, spent spent adopting burnup credits that can be rationally designed without taking excessive safety margins. The present invention relates to a criticality safety design program for fuel transportation casks and spent fuel storage racks.

  When conducting criticality safety design of transport and storage equipment for spent uranium (U) fuel or spent mixed oxide fuel (MOX), conservatively based on new fuel assumptions that do not take into account a decrease in reactivity due to fuel combustion. In many cases, critical safety design was performed. Due to the high enrichment of fuel accompanying the recent high burnup fuel design, the critical safety design based on the above new fuel assumption makes the design overly conservative, leading to high costs for transportation and storage equipment, etc. There is a problem. In order to design reasonable transport and storage equipment, the fissionable nuclide concentration in the fuel assembly is depleted with combustion, producing neutron absorbing materials such as fission products (FP), resulting in the fuel assembly The design using burn-up credits is underway, taking into account the characteristic that the neutron multiplication characteristic of the body decreases.

  However, even in a design employing burnup credits, as shown in "" Burnup Credit Application of Transportation and Storage of Spent BWR Fuel Assemblies ", ICNS'99, Versailles, Vol.4, p.1635 (1999)" In addition, a model bundle that includes all fuels is created without considering the burnup and void ratio of each fuel assembly individually. However, the safety margin is still excessively conservative. In addition, when a fuel different from the assumed spent fuel is loaded, it is necessary to create the model bundle again, which is problematic in terms of versatility.

  Furthermore, as one of the reasons why excessive maintainability must be set, there is almost no criticality test data with spent fuel for demonstrating the design method, and the implementation of the criticality test is very expensive. There is a problem.

  For the criticality safety analysis of spent fuel transportation and storage equipment, the transportation calculation code KENO based on the energy multi-group Monte Carlo method in the SCALE system developed in the world-proven US is often used. The Monte Carlo method is used not only for transport calculations but also in many fields. The principle is to determine, for example, by generating random numbers whether neutrons are absorbed in a certain region and what the energy of neutrons after scattering will be.

  The Monte Carlo method is a method that faithfully simulates a physical phenomenon. Since a geometric system can be freely described, the accuracy of the solution is high, but a lot of calculation time is required. Monte Carlo calculation codes for solving transport calculations of photons such as neutrons or γ rays can be broadly divided into two groups: an energy multi-group Monte Carlo method and a continuous energy Monte Carlo method.

  The energy multi-group Monte Carlo method deals with neutron energy divided into groups. For example, when energy is divided into two groups, the first group is> 1 MeV and the second group is <1 MeV, the neutron energy is There are only two types. On the other hand, since the continuous energy Monte Carlo method does not divide the energy, the energy of neutrons is infinite.

  The energy multi-group Monte Carlo method has the advantage that the calculation time is fast because the number of energy is small, and the continuous energy Monte Carlo method has the advantage that the calculation time is long but the accuracy is high.

  The continuous energy Monte Carlo code includes the MCNP developed at the Los Alamos Laboratory in the US and the MVP developed at the Japan Atomic Energy Research Institute. The KENO code is an energy multi-group Monte Carlo method, which is reported to be fast in calculation speed and sufficiently accurate for spent fuel transportation and storage equipment systems. This system has good accuracy even for systems with large neutron leakage, such as transport cask, but cannot perform combustion calculations for fuel assemblies with complex enrichments such as BWRs. In order to create the nuclear constant of the field, it is necessary to create the above model bundle. In addition, the collision probability indicated by “Development and Validation of TGBLA Lattice Physics Method”, Proc. ANS Topical Mtg. On Reactor Physics and Shielding, Chicago, Illinois, Vol.I, p.364 (1984) The fuel assembly design code based on the method can accurately calculate the combustion of fuel assemblies even for fuel assemblies with complex enrichment distributions. These codes are based on diffusion calculations. Designed to create minority energy group diffusion constants for use in core process computers and 3D core simulation codes to calculate properties. The energy group division of the minority energy group is usually about 1 to 3 groups.

  The collision probability method is an approximate solution of the neutron transport equation. Dividing the system into a finite number of regions, calculating the probability of whether the neutrons in each region are absorbed or escaping outside the system, the calculated probability is called the collision probability, and based on this collision probability, the transport equation The method of solving is called the collision probability method.

  Generally, the core of a nuclear reactor is composed of a non-homogeneous region in which fuel is rod-shaped and moderators or coolants are present around it. The collision probability method is widely used as a method for accurately handling non-homogeneous regions. Diffusion calculations have good accuracy for large cores with low neutron leakage, but for systems with large neutron leakage such as transport cask, the error is higher than the transport calculation code based on the Monte Carlo method described above. May become large.

  The core process computer is a power plant that monitors the neutron flux distribution, water temperature, control rod position, etc. of the operating reactor, records it in the storage device as operation data, and uses the recorded operation data to measure actual values. It is a computer provided with a learning function for correcting the calculated value so as to suit. The model introduced in this computer is almost the same as the model used in the three-dimensional core simulator, but a simple model may be used in part to improve the calculation speed.

The three-dimensional core simulator is a computer program that calculates the nuclear characteristics of the entire core using the diffusion constant of the fuel assembly unit created by the collision probability method. In the three-dimensional core simulator, a calculation method called a node method is used. In a computer program that calculates a multidimensional system, which is usually divided into core nodes by dividing a 15 cm cube of fuel assemblies into 24 equal parts in the axial direction, each node is divided into core nodes. Divide the space into small meshes like a mesh. In general, the larger the number of meshes, the higher the accuracy, but the longer the calculation time. In the node method, in order to improve the accuracy even with a coarse mesh, an ingenuity model such as an analytical model representing the neutron flux distribution in the node is incorporated (reference: published by the Atomic Energy Society of Japan, Current status and prospects ”).
Japanese Patent Laid-Open No. 9-80191 “Burnup Credit Application of Transportation and Storage of Spent BWR Fuel Assemblies”, ICNS'99, Versailles, Vol.4, p.1635 (1999) “Development and Validation of TGBLA Lattice Physics Method”, Proc. ANS Topical Mtg. On Reactor Physics and Shielding, Chicago, Illinois, Vol.I, p.364 (1984)

  The present invention converts a diffusion constant created from a fuel assembly design code based on a collision probability method capable of performing combustion calculation of a fuel assembly with high accuracy into a transport cross section that can be used in an energy multi-group Monte Carlo method. In addition, using the burnup and average void fraction obtained from the core process computer or core simulation code, the nuclear constant of the fuel assembly is created, and energy that can accurately evaluate subcriticality even in systems with large neutron leakage It is an object of the present invention to provide a critical safety design program for a spent fuel transportation cask and spent fuel storage rack, which employs a reasonable burnup credit by the multi-group Monte Carlo method.

  The present invention converts the burnup and void fraction dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method into a transport cross section, and obtained from the core process computer within this transport cross section. By extracting the portion corresponding to the average void fraction that is the average of the burnup of the fuel assembly and the void fraction over the entire combustion period, the transport cross-sectional area of the spent fuel assembly is created and the spent fuel assembly is transported. And calculating the subcriticality of the equipment for transporting and storing the spent fuel assembly by the energy multi-group Monte Carlo method using the storage arrangement and the transport cross-sectional area of the spent fuel assembly. To do.

  Further, the present invention converts the burnup and void ratio dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method into a transport cross section, and among these transport cross sections, a three-dimensional core simulation code By extracting the portion corresponding to the combined void ratio obtained by averaging the burnup of the fuel assembly obtained from the above and the void ratio of the entire combustion period, a transport cross-sectional area of the spent fuel assembly is created, and the spent fuel Calculate the subcriticality of the equipment that transports and stores the spent fuel assembly by the energy multi-group Monte Carlo method using the arrangement for transporting and storing the assembly and the transport cross section of the spent fuel assembly. It is characterized by that.

  According to the present invention, a reasonable burnup credit is adopted by using the fuel design code by the collision probability method, the energy multi-group Monte Carlo method, the burnup obtained from the core process computer or the core simulation code, and the average void fraction. Critical safety design of spent fuel transportation and storage equipment.

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

  FIG. 1 is a block diagram showing a basic configuration example as one embodiment of the present invention. As shown in FIG. 1, the first input file 1 specifying the geometric dimensions of the fuel assembly, the burnup step, the initial fuel composition, the power density, and the void ratio is created for all the fuel types to be used. The fuel assembly combustion calculation is performed with the fuel design code 2 based on the probability method. Then, the fuel assembly diffusion constant (D), total cross section (T), absorption cross section (A), deceleration cross section (SL), and neutron production obtained as a result of the above calculation The cross-sectional area (P) is stored in the first output file 3. On the other hand, in the second input file 4, the fuel assembly arrangement in the core, the control rod insertion position, the power density, the core operation and the shutdown plan are specified. Fuel assembly arrangement, control rod insertion position, power density, core operation and shutdown plan, burnup / void rate dependent fuel assembly diffusion constant (D), total stored in the first output file 3 The cross section (T), absorption cross section (A), deceleration cross section (SL), and neutron production cross section (P) are input to the core process computer or the three-dimensional core simulator 5 where the spent fuel assembly is burned. And the average void rate during combustion are calculated, and the burn rate of the spent fuel assembly and the average void rate during combustion are stored in the second output file 6.

  On the other hand, each constant stored in the first output file 3 is input to the cross-sectional area conversion code 7 by the diagonal transport approximation method. Here, each constant indicated by D, T, A, P, TR, SL, and S is energy-dependent, and precisely D (E), T (E), A (E), P (E), TR (E), SL (E ′ → E), S (E ′ → E). The deceleration cross section SL represents the probability of scattering from an energy group E ′ to the energy group E, but does not include the scattering probability within the energy group E (referred to as self-group scattering cross section). That is, E ≠ E ′. On the other hand, the scattering cross section S (E ′ → E) includes the self-group scattering cross section S (E → E). The deceleration sectional area SL (E ′ → E) when E ≠ E ′ is equal to the scattering sectional area S (E ′ → E). This difference is due to the fact that the self-scattering cross section is not required in the diffusion calculation, but the self-scattering cross section is required in the Monte Carlo calculation.

Transport cross sections are described in general nuclear reactor physics reference books (for example, JJ Doderstadt, Hyundai Engineering, Masakuni Narita and Fumiyuki Fujita, "Reactor theory and analysis" by LJ Hamilton) (Chapter 4 etc.) If the angular dependence is expressed by Legendre expansion,

It becomes. here,
E ': Incident neutron energy group
E: Scattered neutron energy group μ: Scattering cosine Σ _s (E '→ E, μ): Angle-dependent scattering cross section Σ _sl (E' → E): l-order scattering cross section
P _l (μ): l-order Legendre coefficient.

Here, the zero-order and first-order scattering cross sections are as follows.

Here, Σ _s (E ′ → E) is an angle integral of the angle-dependent scattering cross section Σ _s (E ′ → E, μ). μ is the scattering cosine.

With the above scattering cross section, the fuel design code that uses the collision probability method can output only Σ _s (E '→ E), so the angle dependence of the transport cross section cannot be considered. .

Therefore, the total cross-sectional area Σ _t (E) is replaced with the transport cross-sectional area Σ _tr (E) so that a substantially equivalent solution can be obtained when considering substantially the first-order angle dependence. Σ _tr (E) is expressed as follows.

Σ _tr (E) = 1 / 3D (E) = Σ _t (E)-μ (E) Σ _s (E)
Here, D (E) is a diffusion constant of the E group, and is output from the fuel design code.

Σ _s (E) is obtained by integrating Σ _s (E ′ → E) with the incident energy group E ′.

The above operation corrects the effect of the primary component of scattering with the total cross-sectional area. Here, the integral value μ (E) Σ _s (E) of the incident energy group E ′ can be evaluated from the above Σ _tr (E) and Σ _t (E), but the differential value μ (E ′ → E) · Σ _s (E '→ E) cannot be evaluated. Therefore, in order to maintain the neutron balance, the self-group scattering cross section Σ _s (E → E) is replaced with Σ _s (E → E) −μ (E) Σ _s (E).

  The above operation is called a diagonal transport approximation method, and by this operation, it is possible to create a zero-order transport cross-sectional area (P0 cross-sectional area) in consideration of the primary angle component from the fuel design code.

  The newly required cross sections for the Monte Carlo calculation are the transport cross section TR and the self-scattering cross section S (E → E). In the following, the cross-sectional area conversion operation will be described taking the energy 3 group as an example. The numbers in parentheses indicate energy.

TR (1) = 1 / 3D (1)
TR (2) = 1 / 3D (2)
TR (3) = 1 / 3D (3)
S (1 → 1) = T (1) − (A (1) + SL (1 → 2) + SL (1 → 3))
S (2 → 2) = T (2) − (A (2) + SL (2 → 1) + SL (2 → 3))
S (3 → 3) = T (3) − (A (3) + SL (1 → 2) + SL (1 → 3))
These constants are determined depending on the burnup and void. That is, in the first energy group, TR (1, BU (i), V (j)), A (1, BU (i), V (j)), P (1, BU (i), V ( j)),
S (1 → 1, BU (i), V (j)), S (1 → 2, BU (i), V (j)), S (1 → 3, BU (i), V (j)) Are obtained as a table and stored in the third output file 8.

  Here, in order to obtain the transport cross section (TR), absorption cross section (A), neutron production cross section (P), and scattering cross section (S) of a specific spent fuel assembly, the second output file Using the burnup (BU (x)) and void fraction V (y) stored in Fig. 6, the transport cross section (TR), absorption cross section (A), and neutron production interruption of the spent fuel assembly By the code 9 for calculating the area (P) and the scattering cross section (S), the burnup (BU (x)) and the void ratio V (y) are x and y in the table of the third output file 8 Then, the transport cross section (TR), absorption cross section (A), neutron production cross section (P), and scattering cross section (S) at that time are extracted. If not, the transport cross section (TR), absorption cross section (A), neutron production cross section (P), and scattering cross section (S) are calculated by extrapolation. The transport cross section (TR), absorption cross section (A), neutron production cross section (P), and scattering cross section (S) calculated by the above extraction or extrapolation are stored in the fourth output file 10. . That is, calculating the burnup and void ratio of all fuels requires enormous calculation time, so only the typical burnup and void ratio are calculated, and the spent fuel is calculated by interpolation and extrapolation as described above. The calculation time can be shortened by determining the nuclear constant.

  At this time, if it is necessary to save calculation time and storage capacity of the computer, further group by fuel type and the outermost periphery of the core, etc., and the void fraction of the fuel assemblies in the group should be the same. As a result, the number of average void fractions can be reduced, and the calculation time can be further shortened. .

  FIG. 2 illustrates in detail the flow from the creation of the first output file 3 to the fourth output file 10.

  Next, a critical safety analysis of the spent fuel transport cask is performed using the constants stored in the fourth output file 10. That is, the fuel assembly arrangement and the structural material arrangement of the spent fuel transport cask are input to the third input file 11, and other than the nuclear constant of the spent fuel assembly stored in the fourth output file 10. A fourth input file 12 for creating a constant of the structural material is created, and the composition and dimensions of the structural material are input to the fourth input file 12, and one-dimensional is based on the composition and dimensions of the structural material. The structural material constant is calculated by the code 13 for calculating the transport cross section (TR), absorption cross section (A), and scattering cross section (S) by the system. This code is known and can also be performed with the SCALE code. The constants of the structural material have less influence on the critical safety design than the fuel assembly, and sufficient accuracy can be obtained by a constant creation method using an existing method.

  The transport cross section (TR), absorption cross section (A), and scattering cross section (S) of the structural material other than the fuel assembly calculated by the above code 13 are obtained by the Monte Carlo method via the fifth output file 14. This is input to a code 15 for calculating the subcriticality of the spent fuel transport cask. The code 15 for calculating the subcriticality of the spent fuel transport cask by the Monte Carlo method includes information on the fuel assembly layout and structural material layout of the spent fuel transport cask in the third input file 11, and the fourth The transport cross section (TR), absorption cross section (A), neutron production cross section (P), and scattering cross section (S) stored in the output file 10 are also input. The subcriticality of the cask is calculated and the uncorrected subcriticality of the spent fuel transport cask is stored in the sixth output file 16.

  By the way, the subcriticality stored in the sixth output file 16 is an uncorrected subcriticality evaluated only by calculation as described above. Therefore, the ratio between the measured value and the calculated value obtained from the results of the critical experiment analysis and the actual machine cold / hot critical test analysis is created in the fifth input file 17 as a correction factor. Thus, in the code 18 for correcting the calculated subcriticality, the uncorrected subcriticality stored in the sixth output file 16 is corrected by the correction factor created in the fifth input file 17. Finally, the corrected subcriticality of the spent fuel transport cask can be obtained in the seventh output file 19.

  Considering the statistical error 3σ of the Monte Carlo calculation after the correction, the design is completed if it is less than the limit value, and redesign is performed if it exceeds the limit value.

  Calculation codes may have systematic errors due to calculation methods and data. Therefore, the systematic error was evaluated from the results of criticality experiment analysis and actual cold / hot criticality test analysis using several criticality test devices simulating the target fuel assembly, and the spent fuel transportation and storage equipment calculated was calculated. The design accuracy can be improved by correcting the subcriticality.

  The actual cold temperature criticality is a state where no power is output when the nuclear reactor is started up, that is, a critical state where the water temperature is normal temperature.

  When performing a cold / cool criticality test analysis using a conventional critical safety design method, it is necessary to set the nuclide composition for each fuel rod in the fuel assembly, taking into account the void fraction and power density. Data entry was necessary and practically impossible. The fuel design code based on the present invention is originally intended for core design, and the cross-sectional area for critical safety design obtained after cross-sectional area conversion is also the power density, void ratio, and composition of each fuel rod. Since all of them have been taken into consideration, a huge amount of actual device cold / hot criticality test data can be used as demonstration data, and critical test data of spent fuel can be obtained at low cost. In addition, in the conventional method, there was no verification data, so excessive maintainability had to be set.However, the existence of verification data confirmed the accuracy of the method, and the difference between the test value and the calculated value. From this, the calculated value can be corrected.

  FIG. 3 is a block diagram showing a basic configuration as another embodiment of the present invention, and several fission products having negative reactivity in spent fuel in order to improve the maintainability of the design. And actinides are excluded. That is, the cross-sectional area of a specific fission product or actinide having negative reactivity in the spent fuel among the constants of the first output file 3 is set to 0, and the cross-sectional area of the specific fission product or actinide is set to 0. Fuel assembly diffusion constant (D), total cross section (T), absorption cross section (A), removal cross section (R), and neutron production cross section (R) P) is stored in the eighth output file 21. The constant storage format of the eighth output file 21 is the same as that of the first output file 3, and the rest is the same as in FIG. The burnup information from the core process computer or the three-dimensional core simulator is not used.

  Thus, in the present embodiment, those having a negative reactivity are excluded, and the resulting subcriticality is higher than the subcriticality in the first embodiment, and is close to the criticality. Although it is disadvantageous in terms of design, maintainability is high. This is particularly effective when the composition error of the fuel assembly after combustion seems to be large, or when there is little data for critical experiment analysis or actual equipment cold / warm test analysis and the error of the correction factor seems to be large.

  FIG. 4 is a block diagram showing an example of the basic configuration of another embodiment of the present invention. In order to further improve maintainability, the combustion of the fuel assembly is not considered and the neutrons arranged in the fuel assembly are not considered. It is a so-called “new fuel assumption” that does not consider gadolinia, which is a strong absorber, and is the most maintainable method.

  That is, since the combustion calculation is not performed in the present embodiment, only the geometric composition and the initial fuel composition of the fuel assembly are input in the first input file 1. Gadolinia is excluded from the initial fuel composition. Therefore, an unburned constant that does not include gadolinia is calculated by the fuel design code 2 based on the collision probability method and stored in the ninth output file 23. Each constant stored in the ninth output file 23 is a diagonal line. The cross section conversion code 7 is input by the transport approximation method. The transport cross section (TR), absorption cross section (A), neutron production cross section (P), and scattering cross section (S) of the fuel assembly calculated here are input. Are stored in the tenth output file 24. The transport cross section (TR), absorption cross section (A), neutron production cross section (P), and scattering cross section (S) of the above fuel assembly indicate the subcriticality of the spent fuel transport cask by the Monte Carlo method. The code 15 to be calculated is input. The following operations can be performed in the same manner as in FIG.

  FIG. 5 is a block diagram showing a further embodiment of the present invention. The system is more streamlined than the one shown in FIG. 4, and is a system called Gadori credit. That is, a fuel assembly having the highest reactivity is selected from the fuel assembly constants depending on the burnup, power density, and void ratio obtained in the first output file 3, and the fuel assembly is selected. Are stored in the eleventh output file 25, and the constants of this fuel assembly, that is, diffusion constant (D), total cross section (T), absorption cross section (A), deceleration cross section (SL), and neutron generation The cross-sectional area (P) is converted by the cross-sectional area conversion code 7 by the diagonal transport approximation method. The following operations can be performed in the same manner as in FIGS.

The BWR fuel assembly is loaded with gadolinia in order to suppress the reactivity at the initial stage of combustion. Since gadolinia is a strong neutron absorber, it decreases with combustion. Accordingly, the gadolinia decreases and the reactivity increases with combustion, but the fuel itself also decreases due to the combustion. Therefore, the reactivity decreases with a certain degree of combustion as a peak. Therefore, criticality safety analysis is performed using the peak value of the reactivity as the reactivity of the spent fuel assembly. Therefore, since this peak value is smaller than the new fuel assumption shown in FIG. 4, a more rational design than the program shown in FIG. 4 can be performed.

  As described above, according to the present invention, the fuel design code based on the collision probability method, the energy multi-group Monte Carlo method, the burnup and the average void ratio obtained from the core process computer or the three-dimensional core simulator can be used. It is possible to carry out critical safety design of transport and storage containers for spent fuels using various burnup credits.

  Also, by converting the burnup and void fraction dependent nuclear constants into a transport cross section using diagonal transport approximation, a transport cross section for Monte Carlo calculation with high accuracy can be created. Furthermore, it can be obtained when subcriticality obtained when analyzing some test results from some critical experimental equipment, or when analyzing some test results from cold criticality tests of some nuclear power plants. With the correction value evaluated from the subcriticality, the accuracy of the subcriticality of the spent fuel transportation and storage equipment can be improved.

  Further, when determining the burnup and void ratio dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method, the burnup and void ratio are calculated at appropriate intervals, and the core process computer or For the nuclear constant corresponding to the burnup and average void fraction of the fuel assembly obtained from the core simulation code, the burnup and void fraction dependent nuclear constant calculated at the appropriate intervals are interpolated or extrapolated. Thus, the calculation time can be shortened.

  Further, the average void fraction of the fuel assembly obtained from the core process computer or the core simulation code is classified for each fuel type, and the classified fuel type is further classified into the outermost periphery of the core and others, and the separated fuel For a group, the number of calculation of the void rate dependent nuclear constant can be shortened by adopting the average void rate of the fuel group.

  Furthermore, the calculation accuracy can be improved by using a nuclear thermal coupling calculation method based on the multi-group modern node method as the core process computer or the core simulation code.

  Further, by using a substance sensitive to neutrons, a scintillator, and an optical fiber, a small or plate-like neutron detector can be configured, and measurement can be performed efficiently even in a narrow space. Furthermore, when calculating the burnup and void fraction dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method, the neutron absorption effect by some fission products or the neutron absorption effect by minor actinides By excluding or both of them, a critical safety design with a higher safety margin can be performed.

  Further, for the structural material other than the fuel assembly of the spent fuel transportation and storage equipment, and water, the accuracy regarding the anisotropy of neutron scattering can be improved by using the transportation cross section calculated by the transportation calculation. Here, transport calculation solves neutron behavior using neutron position, neutron energy, and neutron direction as parameters.

  The above calculation usually requires a lot of calculation time and memory on the computer. In the design of the reactor core, in order to improve calculation efficiency, it is assumed that neutrons are uniformly scattered in all directions, and calculation is performed while ignoring information on the direction of neutrons. This approximation is called diffusion approximation, and this calculation is called diffusion calculation for transport calculation. This diffusion calculation has a large error where the isotropic scattering does not hold. The deviation from isotropic scattering is called neutron scattering anisotropy, and the anisotropy increases at boundaries where matter changes. In the spent fuel transportation and storage equipment, which is the subject of this patent, anisotropy increases at the boundary between the fuel and the structural material, the boundary between water and air, and the like. In this patent, the transport cross-sectional area of the fuel assembly part is created by a diagonal transport approximation of a constant prepared for diffusion calculation, and includes a factor that causes an error in the information related to the neutron direction. Since the construction of the transport cross section of the fuel assembly is complicated, it is difficult to handle homogeneously, and it is difficult to solve it by transport calculations. In this patent, the transport cross section is created using the fuel design code. doing. On the other hand, the structural material part and water can be handled homogeneously from the viewpoint of neutron behavior, and can be easily solved using a public program such as the one-dimensional transport calculation program XSDRN code in the SCALE system described above. be able to. Therefore, although the anisotropy of neutrons increases, for structural materials and water that can be handled homogeneously, a transport cross section is created by transport calculation to improve the accuracy of neutron scattering anisotropy.

  Furthermore, even in the design using new fuel assumptions and Gadori Credit, rational critical safety design can be performed, and by setting a conservative irradiation history, critical safety design with a higher safety margin can be achieved. It can be carried out.

The conceptual diagram for demonstrating the critical-safety design program of the spent fuel transportation and storage container by embodiment of this invention. The conceptual diagram of the second half of the criticality safety design program shown in FIG. The conceptual diagram which shows other embodiment of this invention. The conceptual diagram which shows other embodiment of this invention. The conceptual diagram which shows other embodiment of this invention. The conceptual diagram of the flow until producing a 4th output file from a 1st output file.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 1st input file 2 Fuel design code by collision probability method 3 1st output file 4 2nd input file 5 Core process computer or core simulation code 6 2nd output file 7 Cross section conversion code by diagonal transport approximation method 8 Third output file 9 Code 10 for calculating the transport cross section TR of the spent fuel assembly, etc. Fourth output file 11 Third input file 12 Fourth input file 13 Transport cross section, absorption in a one-dimensional system Code 14 for calculating the cross-sectional area and scattering cross-section 14 Fifth output file 15 Code 16 for calculating the subcriticality of the spent fuel transport cask by the Monte Carlo method Sixth output file 17 Fifth input file 18 Code 19 to correct criticality 7th output file 20 Cross section of a specific fission product, actinide The output file 25 of the cord 21 eighth output file 23 ninth output file 24 a tenth of that 0 eleventh output file

Claims (14)

  The burnup and void fraction dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method is converted into a transport cross section, and within this transport cross section, the fuel assembly obtained from the core process computer is converted. By extracting the portion corresponding to the average void fraction obtained by averaging the burnup and the void fraction of the entire combustion period, the transport cross section of the spent fuel assembly is created, and the spent fuel assembly is transported and stored. The subcriticality of the equipment that transports and stores the spent fuel assemblies is calculated by the energy multi-group Monte Carlo method using the arrangement of the above and the transport cross section of the spent fuel assemblies. Criticality safety design program for fuel transportation and storage equipment.   The fuel assembly burnup and void fraction dependent nuclear constants calculated by the fuel design code based on the collision probability method are converted into the transport cross section, and the fuel set obtained from the 3D core simulation code within this transport cross section By extracting the portion corresponding to the average void fraction that is the average of the burnup of the body and the void fraction over the entire combustion period, the transport cross section of the spent fuel assembly is created, and the spent fuel assembly is transported and stored. The subcriticality of the equipment that transports and stores the spent fuel assemblies is calculated by the energy multi-group Monte Carlo method using the arrangement and the transport cross-sectional area of the spent fuel assemblies. Critical safety design program for spent fuel transport and storage equipment.   The criticality of spent fuel transport and storage equipment according to claim 1 or 2, characterized in that diagonal transport approximation is used as means for converting the burnup and void fraction dependent nuclear constant of the fuel assembly into a cross section for transport. Safety design program.   To correct the subcriticality of the equipment that transports and stores the spent fuel assemblies, using the ratio between the subcriticality obtained when analyzing the test results from several critical experimental devices and the calculated value as a correction factor. The critical safety design program for spent fuel transportation and storage equipment according to claim 1 or 2.   Subcriticality of equipment that transports and stores spent fuel assemblies, with the ratio between the subcriticality obtained when analyzing the results of cold criticality tests at several nuclear power plants and the calculated value as a correction factor The critical safety design program for spent fuel transportation and storage equipment according to claim 1 or 2, characterized in that   When determining the burnup and void ratio dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method, the burnup and void ratio are calculated at appropriate intervals, and the core process computer or three-dimensional The calculation time is shortened by calculating by interpolating or extrapolating the nuclear constant corresponding to the burnup and average void ratio of the fuel assembly obtained from the core simulation code. 4. A critical safety design program for spent fuel transportation and storage equipment according to any one of 3 above.   The average void fraction of the fuel assembly obtained from the core process computer or the three-dimensional core simulation code is classified for each fuel type, and the classified fuel type is further classified into the outermost periphery of the core and others, and the classified fuel 3. The spent fuel transportation and storage device according to claim 1 or 2, characterized in that the number of calculations of the void rate dependent nuclear constant is shortened by adopting an average void rate of the fuel group. Criticality safety design program.   The spent fuel transportation and storage according to claim 1 or 2, wherein calculation accuracy is improved by using a nuclear thermal coupling calculation based on a multi-group modern node method as the core process computer or the three-dimensional core simulation code. Equipment criticality safety design program.   When calculating the burnup and void fraction dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method, the safety margin can be increased by excluding the neutron absorption effect by some fission products. The criticality safety design program for spent fuel transportation and storage equipment according to claim 1 or 2, characterized in that a subcriticality with an increased value is calculated.   When calculating the burnup and void fraction dependent nuclear constant of the fuel assembly calculated by the fuel design code based on the collision probability method, the neutron absorption effect due to some minor actinides is excluded, thereby increasing the safety margin. The critical safety design program for spent fuel transportation and storage equipment according to claim 1 or 2, wherein the increased subcriticality is calculated.   The structural material other than the fuel assembly of the spent fuel transport and storage equipment, and water, the accuracy of neutron scattering anisotropy is improved by using the transport cross section calculated by transport calculation, Item 3. A critical safety design program for spent fuel transportation and storage equipment according to item 1 or 2.   Create a transport cross-section of the fuel assembly converted from the nuclear constant calculated for the unburned fuel assembly that is assumed not to contain Gd in the fuel design code based on the collision probability method. The subcriticality of the equipment that transports and stores the spent fuel assembly is calculated by the energy multi-group Monte Carlo method using the transportation and storage arrangement and the transport cross section of the fuel assembly. A critical safety design program for spent fuel transportation and storage equipment according to claim 1 or 2.   The spent fuel assembly is calculated by using the transport cross section of the fuel assembly, which is converted from the nuclear constant of the burnup that is most reactive throughout the combustion period, calculated by the fuel design code based on the collision probability method. The subcriticality of the equipment that transports and stores the spent fuel assembly is calculated by the energy multi-group Monte Carlo method using the arrangement for transporting and storing the body and the transport cross section of the fuel assembly. The critical safety design program for spent fuel transportation and storage equipment according to claim 1 or 2.   A fuel design code based on the collision probability method, and the transport interruption converted from the nuclear constant of the fuel assembly calculated using the power density and average void fraction selected to achieve the highest reactivity within the same burnup. Transportation and storage of spent fuel assemblies by an energy multi-group Monte Carlo method using an arrangement for transporting and storing the spent fuel assemblies and the cross-sectional area of transport of the spent fuel assemblies. The criticality safety design program for spent fuel transportation and storage equipment according to claim 1, wherein the subcriticality of the equipment to be used is calculated.

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