A method and device for calculating the accumulation of fission products in the gap between fuel cladding and fuel cladding

Abstract

A method for calculating the fission product accumulation in the fuel-cladding gap calculates the in-core accumulation and production rate of target nuclides based on the reactor's fuel cycle management parameters, fuel assembly design parameters, and primary system design parameters. The power distribution and burnup distribution within the fuel rods are then calculated. Based on the fuel rod burnup and temperature distribution, the release and production ratios of different target nuclides from the pellets to the cladding gap are calculated. This method then determines the temporal accumulation of various fission products in the fuel-cladding gap under operating conditions. This method accurately simulates and calculates the fission product accumulation in the fuel-cladding gap, effectively guiding reactor design and safety management. The present invention also provides a calculation device.

Description

Technical Field

[0001] The present invention belongs to the field of nuclear power, and in particular relates to a method and a device for calculating the accumulation of fission products in the gap between fuel claddings. Background Art

[0002] During nuclear reactor operation, fuel fission reactions generate a large number of radioactive fission products. These fission products are released from the fuel pellets through diffusion, recoil, and ejection mechanisms and accumulate in the cladding gap between the fuel rods and the fuel pellets. Fission products in the cladding gap are a significant source of radioactive material generated during nuclear power plant operation, including radioactive isotopes of various elements, such as noble gases (Xe and Kr), halogens (I and Br), and alkali metals (Cs and Rb). Under accident conditions, leakage of radioactive products from the cladding gap is a significant factor leading to radiological consequences both within and outside the accident site. Therefore, accurately calculating and assessing the fission product inventory in the cladding gap is crucial for nuclear power plant design, safety management, and assessment of radiological consequences of accidents. Currently, the estimation of fission product accumulation in the fuel cladding gap is primarily based on engineering assumptions based on RG1.183-2000 and NB / T20444-2017RK. This estimation method fails to reflect the design and structural differences between different reactor types. Furthermore, due to its conservative values, it fails to reflect actual production, making its guidance for engineering design limited. Furthermore, this assessment method fails to distinguish between the differences of other nuclides besides I-131 and Kr-85, failing to accurately reflect actual conditions. Therefore, providing a calculation method for the fission product accumulation in the fuel cladding gap, through refined calculation of quantitative accident source terms and assessment of accident radiological consequences, has positive implications for improving the rationality of nuclear power plant system design and dedicated safety features, as well as the economic viability of small reactor development. Summary of the Invention

[0003] The present invention aims to provide a method for calculating the amount of fission product accumulation in the fuel cladding gap, which quantitatively calculates the amount of fission product accumulation in the cladding gap. The present invention also provides a calculation device.

[0004] According to an embodiment of one aspect of the present invention, a method for calculating the fission product accumulation in the gap between fuel rod claddings is provided, the method comprising the following steps:

[0005] a) Provide the fuel cycle management parameters, fuel assembly design parameters and primary system design parameters of the target reactor;

[0006] b) providing an ignition consumption calculation model, calculating the core accumulation of the target nuclide i over time based on the parameters provided in step a), and further calculating the generation rate B of the target nuclide i , where B is the generation rate;

[0007] c) Calculate the power distribution p(r, z) and burnup distribution Bu(r, z, t) within the fuel rod, where p is the power, Bu is the burnup distribution, r is the radial position of the calculation node, z is the axial position of the calculation node, and t is time;

[0008] d) Calculate the radial and axial temperature distribution inside the fuel rod based on p(r,z);

[0009] e) Calculate the release ratio (R / B) of the target nuclide from the pellet to the cladding gap by combining the burnup distribution and the power distribution. i , where R is the number of atoms released in a given space per unit time;

[0010] f) Binding of different target nuclides (R / B) i The equilibrium relationship of the fission product accumulation in the fuel cladding gap is calculated to obtain the fission product accumulation in the fuel cladding gap over time under operating conditions.

[0011] By using the above method, the accumulation of different types of nuclides generated during the service of the fuel rods can be quantitatively calculated respectively, thereby obtaining accurate calculation results of the total amount of fission product accumulation and improving the accuracy of simulation calculations.

[0012] Furthermore, in some embodiments, in step a), the fuel cycle management parameters include fuel enrichment, power density, irradiation cycle length, and core axial power distribution; the fuel assembly design parameters include material composition and fuel rod size parameters; and the primary loop system design parameters include pressure, temperature, and flow rate.

[0013] Furthermore, in some embodiments, in the step b), the production rate Bi of the target nuclide i is

[0014] where N i (t) is the number of atoms, is the rate at which the target nuclide i is produced by fission of the fissionable nuclide j, is the rate at which the target nuclide i is produced by the activation reaction of nuclide k with neutrons, is the production rate of the target nuclide i by the radioactive decay of the nuclide m, is the radiation capture extinction rate of the target nuclide i.

[0015] Furthermore, in some embodiments, it is characterized in that the ignition consumption calculation model includes an ignition consumption parameter database, and the ignition consumption parameter database includes reaction cross-section parameters, fission yields, and decay parameters.

[0016] Furthermore, in some embodiments, in step c), the power distribution includes radial power distribution and axial power distribution, wherein the radial power distribution g(r)=C×I0(κ×r), where g(r) is the radial power distribution function of the fuel rod, C is a normalization constant, and I0(x) is a first-kind zero-order modified Bessel function; Where ε is the fuel enrichment percentage, ρ U is the U density in the fuel rod, R fuel is the radius of the fuel pellet; the axial power distribution f(z) adopts the axial power distribution of the fuel rod in the reactor design, or adopts Where L is the length of the fuel rod p(r,z) = C×P rod ×g(r)×f(z), where Prod is the average power of the fuel rod; where ΔV r,z is the fuel volume at the (r,z) node.

[0017] Furthermore, in some embodiments, in step d), the temperature distribution of the fuel rod is calculated using the following model:

[0018]

[0019]

[0020]

[0021]

[0022]

[0023] Among them, T b is the mainstream temperature of the single-channel coolant, T in is the inlet temperature of a single channel, T cs is the outer surface temperature of the cladding, T ci is the inner surface temperature of the cladding, T fs is the fuel pellet surface temperature, T p is the temperature distribution inside the fuel pellet, D0 is the equivalent diameter of a single channel, A f is the cross-sectional area of ​​a single channel, q n is the radial heat flux of the fuel rod, G is the mass flux of the coolant in a single channel, C p is the constant pressure heat capacity of the coolant, h f is the convective heat transfer coefficient, q i is the radial linear power density of the fuel rod, r o is the outer diameter of the fuel rod cladding, r i is the inner diameter of the fuel rod cladding, k c is the thermal conductivity of the fuel rod cladding, kgas is the thermal conductivity of the gas in the cladding gap, q v is the volume heat release rate of the fuel rod, k fuel is the thermal conductivity of the fuel pellet.

[0024] Furthermore, in some embodiments, in step e), R / B is calculated by solving the equation We can obtain , where C is the nuclide concentration in the fuel, λ is the nuclide decay constant, and D is the nuclide diffusion coefficient.

[0025] Furthermore, in some embodiments, for the target nuclide i,

[0026]

[0027] in, is the release production ratio of nuclide i at the node position (r, z); is the ratio of the surface area to the volume at the node position (r, z); F i is the conversion factor for long-lived nuclides, H j,i is the correction factor for the diffusion capacity of parent nucleus j to nuclide i; τ = D × t / a 2 , where D is the diffusion coefficient of the nuclide and a is the radius of the ideal sphere equivalent to the fuel pellet grain.

[0028] Furthermore, in some embodiments, D=D1(T)+D2(T,F)+D3(F), where D1 is the intrinsic diffusion coefficient, D2 is the diffusion coefficient of the vacancy mechanism increased by irradiation, D3 is the non-thermal diffusion, T is the fuel pellet temperature, and F is the fission rate.

[0029] Furthermore, in some embodiments, There is a sudden change in temperature T caused by the combined effect of bubbles link , in the calculation, the temperature is greater than T link and less than T link hour Different values ​​are used respectively, among which:

[0030]

[0031] Furthermore, in some embodiments, in step f), the equilibrium relationship of the fission product accumulation in the fuel cladding gap is calculated by the following differential equation:

[0032]

[0033] in,

[0034] Where P is power, V is volume, rod is the fuel rod as a whole, node is the axial node, ring is the radial ring, and A is the g,i (t) is the radioactive accumulation of nuclide i, A g,i (t) is the radioactive inventory of parent nucleus j of nuclide i in the fuel cladding.

[0035] According to an embodiment of another aspect of the present invention, a computing device is provided, which includes a memory and a processor, wherein the memory stores a computing program, and when the computing program is executed by the processor, it is capable of implementing the method for calculating the accumulation of fission products in the gap between the fuel claddings provided in any of the aforementioned embodiments. BRIEF DESCRIPTION OF THE DRAWINGS

[0036] Figure 1 A flow chart for calculating the fission product inventory in the fuel cladding gap in one embodiment;

[0037] Figure 2 A schematic diagram of fuel pellet calculation node division in one embodiment;

[0038] Figure 3 is a curve of nuclear element inventory in a fuel rod pellet in one embodiment;

[0039] Figure 4 is a curve of nucleus generation rate in a fuel rod pellet in one embodiment;

[0040] Figure 5 Schematic diagram of temperature distribution inside a fuel rod in one embodiment;

[0041] Figure 6 This is a curve of fission product accumulation in the cladding gap in one embodiment.

[0042] The purpose of the above drawings is to provide a detailed description of the present invention so that those skilled in the art can understand the technical concept of the present invention, and is not intended to limit the present invention. DETAILED DESCRIPTION

[0043] The present invention will be further described in detail below through specific embodiments with reference to the accompanying drawings.

[0044] Reference to "embodiments" herein means that the specific features, structures, or characteristics described in conjunction with the embodiments may be included in at least one embodiment herein. The phrases appearing in various locations in the specification do not necessarily refer to the same embodiment, nor are they limited to mutually exclusive independent or alternative embodiments. It should be understood by those skilled in the art that the embodiments herein may be combined with other embodiments without causing structural conflicts. In the description herein, "a plurality" means at least two.

[0045] During reactor operation, fuel fission produces a large number of radioactive fission products. Under high temperatures, these fission products escape from the interior of the fuel pellets through diffusion, recoil, and ejection mechanisms, accumulating in the gap between the fuel rod cladding and the fuel pellets. These fission products are highly radioactive. Under normal circumstances, the fission products are contained within the fuel rod cladding and prevent contact with the primary coolant. However, if the cladding is damaged due to defects or accidents, the fission products in the fuel-cladding gap can leak and spread with the coolant, causing serious contamination. In nuclear power plant accident radiological consequence assessments, fission products in the fuel-cladding gap are considered a significant source of radioactivity in events such as loss of coolant accidents (LOCAs), main steam line ruptures, steam generator heat transfer unit ruptures, rod ejection accidents, main pump rotor jams / shaft breakages, and fuel handling accidents. The accumulation of fission products directly affects the on-site and off-site radiological consequences of accidents.

[0046] Currently, according to RG1.183-2000 and NB / T 20444-2017RK, the proportion of inert gases (Xe, Kr), halogens (I, Br), and alkali metals (Cs, Rb) in the interstitial space between fuel pellets and cladding in a LOCA accident is assumed to be 5% of the corresponding core inventory. In contrast, the proportions of interstitial fission products in a non-LOCA accident are: I-131 (8%), Kr-85 (10%), other inert gases (5%), other halogens (5%), and alkali metals (12%). Currently, engineering practice primarily uses these assumptions as the basis for nuclear power plant design and safety assessment. However, with the development of nuclear power technology, the above data cannot reflect the design differences between different pressurized water reactors. From the perspective of safety assessment, the above values ​​are conservative estimates and lack guidance for the technical optimization of nuclear power plant construction plans. In the development of new reactors, especially small reactors, they may deviate from the actual engineering situation. Furthermore, the above values ​​only consider I-131 and Kr-85 in non-LOCA accidents and do not further distinguish the differences of other nuclides. The interstitial shares of different nuclides will vary significantly due to differences in half-life, parent nucleus influence, diffusion capacity, etc.

[0047] To address the above-mentioned issues, an embodiment of the present invention provides a method for calculating the fission product accumulation in the fuel cladding gap, which accurately calculates the changes in the fission product accumulation in the fuel cladding gap over time, thereby providing effective support for quantifying accident source terms, reasonably evaluating the radioactive consequences of accidents, and verifying the rationality of system design and dedicated safety facilities during engineering design, especially small reactor construction.

[0048] In one embodiment, the steps of the method are as follows Figure 1 As shown, the following steps are included:

[0049] Step a): Determine the various reactor parameters to be calculated, including fuel cycle management parameters, fuel assembly design parameters, and primary system design parameters. Specifically, in a preferred embodiment, fuel cycle management parameters include fuel enrichment, power density, irradiation cycle length, and core axial power distribution; fuel assembly design parameters include material composition and fuel rod dimensions; and primary system design parameters include pressure, temperature, and flow rate.

[0050] Step b): Providing an ignition consumption calculation model. For example, an existing reactor ignition consumption calculation program can be used to calculate the core inventory of each target nuclide over time based on the reactor parameters provided in step a), and then processed using an interface program to obtain the fission product production rate.

[0051] In a preferred embodiment, the ignition consumption program is specifically used to calculate the atomic number N of fission products (including actinide nuclides) as a function of time. i (t), defines the net production rate B of fission product nuclide i i : in, is the rate at which fissionable nuclide j produces nuclide i; is the rate at which nuclide i is produced by the activation reaction of nuclide k with neutrons; is the production rate of nuclide i due to the radioactive decay of nuclide m; is the radiation capture extinction rate of nuclide i.

[0052] Furthermore, in a preferred embodiment, the ignition consumption calculation model completes the simulation calculation through the ignition consumption parameter database of the ignition consumption program, and the parameter database includes reaction cross section parameters, fission yield, decay parameters, etc., and the interface program completes B i Calculation of (t).

[0053] Step c): determining radial and axial power distribution and burnup distribution within the fuel rod.

[0054] The radial power distribution is calculated by taking into account the fuel rod self-shielding effect and using the zero-order modified Bessel function: g(r) = C × I0(κ × r). Where g(r) is the fuel rod radial power distribution function; C is a dimensionless normalization constant; I0(x) is the first-kind zero-order modified Bessel function;

[0055]

[0056] In the preferred embodiment, an empirical formula is used Calculation; ε is the fuel enrichment percentage; ρ U is the density of U in the fuel rod (kg·m -3 );R fuelis the radius of the fuel pellet (m); r is the radial position of the calculation stage (m).

[0057] The axial power distribution is preferentially based on the fuel rod axial power distribution in the reactor design scheme, and then the axial power density distribution function is obtained by normalizing the linear power density of each grid. If the axial power distribution of the active area is not available, the fuel rod axial power distribution function can be determined according to the sinusoidal distribution. The power distribution of the fuel rod can be written as p(r,z)=C×P rod ×g(r)×f(z), where p(r,z) is the fuel rod power distribution function, P rod is the average power of the fuel rod (kW).

[0058] Combined with the fuel rod fuel power distribution function, the fuel rod burnup distribution is calculated:

[0059]

[0060] Where Bu(r,z,t) is the fuel consumption of the (r,z) computing node at time t (GWd·tU -1 );ΔV r,z is the fuel volume at the (r,z) node (cm 3 ).

[0061] Step d): Determine the radial and axial temperature distributions inside the fuel rod.

[0062] The temperature distribution within the fuel rod can be calculated using a fuel rod performance analysis program such as FRAPCON, or using a single-channel model. The single-channel model calculation process is as follows: Based on the fuel rod power distribution obtained in step c), the fuel rod temperature distribution is calculated. Due to the large aspect ratio of the fuel rod, axial heat transfer is ignored, and only radial heat transfer is considered. The specific calculation model is as follows:

[0063]

[0064] Among them, T b Mainstream temperature of single-channel coolant (K), T in is the inlet temperature of a single channel (K), T cs is the outer surface temperature of the cladding (K), T ci is the inner surface temperature of the cladding (K), T fs is the fuel pellet surface temperature (K), T p is the temperature distribution inside the fuel pellet (K), D0 is the equivalent diameter of a single channel (m), A f is the cross-sectional area of ​​a single channel (m 2 ), q nis the radial heat flux of the fuel rod (W / m 2 ), G is the mass flux of coolant in a single channel (kg·m -2 ·K -1 ), C p is the constant pressure heat capacity of the coolant (J·kg -1 ·K -1 ), h f is the convective heat transfer coefficient (W·m -2 ·K -1 ), q i is the radial linear power density of the fuel rod (W·m -1 ), r o is the outer diameter of the fuel rod cladding (cm), r i is the inner diameter of the fuel rod cladding (cm), k c is the thermal conductivity of the fuel rod cladding (W·m -1 ·K), k gas is the thermal conductivity of the gas in the cladding gap (W·m -1 ·K),q v is the volume heat release rate of the fuel rod (W·m -3 ), k fuel is the thermal conductivity of the fuel pellet (W·m -1 ·K).

[0065] Step e): Calculate the release ratio R / B of fission products from the pellet to the cladding gap.

[0066] The rate of change of the fission product concentration in the fuel grains described by the ideal diffusion model is equal to the production minus the decay rate plus the diffusion term: Where C is the nuclide concentration in the fuel (n·cm -3 ); r is the radial distance from the center of the equivalent sphere of the fuel grain (cm); D is the diffusion coefficient of the nuclide (cm 2 ·s -1 ); a is the radius of the equivalent sphere of the fuel grain (cm). To solve this equation, the release ratio R / B can be used to characterize the release of fission products, where R is the number of atoms released at a certain position per unit time (n·cm -3 ·s -1 ).

[0067] According to the difference in half-life of nuclides, the calculation model of radionuclide release can be expressed as:

[0068]

[0069] in, is the release production ratio of nuclide i at the node position (r, z); is the ratio of surface area to volume at the node position (r, z) (cm-1 );F i is the conversion factor for long-lived nuclides, H j,i is the correction factor for the diffusion capacity of parent nucleus j to nuclide i; τ = D × t / a 2 .

[0070] In the R / B calculation process, the determination of the diffusion coefficient of fission products in the fuel pellet is a key step.

[0071]

[0072] Where D1 is the intrinsic diffusion coefficient (cm 2 ·s -1 ); D2 is the diffusion coefficient of the vacancy mechanism increased by irradiation (cm 2 ·s -1 ); D3 is the self-diffusion coefficient caused by non-thermal diffusion, (cm 2 ·s -1 ); T is the fuel pellet temperature (K), F is the fission rate (n·cm -3 ·s -1 The diffusion coefficients between different elements can be expressed as: D(Xe)=D(Kr)=D(I)=D(Br) / 20=D(Te) / 4=2D(Cs).

[0073] In the preferred embodiment, the R / B calculation process needs to take into account the impact of S / V changes, that is, the impact of the bubble joint effect on the increase in fission product release. The critical temperature T of the fuel at a certain position is determined by the current burnup size. link , when the temperature at a certain location in the fuel rod is higher than T link When bubbles coalesce, S / V=650cm -1 Otherwise S / V=120cm -1 The critical temperature is expressed as:

[0074]

[0075] By performing comprehensive calculations on all nodes within the fuel rod, the release and production ratio of a given nuclide at constant temperature and burnup can be obtained:

[0076]

[0077] Where P is power (kW·cm -3 ), V is the volume (cm 3 ), rod is the fuel rod as a whole, node represents the axial node, and ring represents the radial ring.

[0078] Step f): Calculate the fission product inventory in the cladding gap.

[0079] Based on the release and production ratios R / B of different nuclides obtained through the above steps, the equilibrium relationship of the fission product accumulation in the fuel cladding gap can be expressed by the following differential equation:

[0080]

[0081] Among them, A g,i (t) is the radioactive accumulation of nuclide i (Bq), A g,i (t) is the radioactive inventory (Bq) of parent nucleus j of nuclide i in the fuel cladding.

[0082] Therefore, numerical analysis methods can be used to obtain quantitative calculation results of the change of fission product accumulation in the fuel cladding gap over time under normal operating conditions, thereby providing guidance for nuclear power plant design, safety management and accident consequence analysis and prediction.

[0083] Another embodiment of the present invention provides a computing device comprising a memory and a processor. The memory stores a program for calculating the fission product inventory in the fuel-cladding gap. When the program is executed by the processor, the method for calculating the fission product inventory in the fuel-cladding gap provided in any of the aforementioned embodiments is implemented. In various embodiments, the computing device can be configured as a general-purpose computer, a dedicated computing device, a dedicated computing server, a virtual machine, or a cloud computing device.

[0084] In a preferred embodiment, the fission product accumulation in the interstitial space of the fuel cladding in the target reactor is simulated and calculated. The target reactor is irradiated at a power density of 40 MW / t for three cycles (500 days per cycle, for a total of 1500 days), with a fuel enrichment of 4%; the outer diameter of the fuel rod cladding is 1 cm and the inner diameter is 0.9 cm; the outer diameter of the fuel pellet is 0.8 cm and the density is 10.95 g / cm 3 The active zone height is 400 cm, the core axial power distribution is sinusoidal; the primary circuit system design pressure is 15.5 MPa, the core inlet temperature is 280 ° C, the core outlet temperature is 320 ° C, and the flow rate is 25m 3 / s. For stacked fuel pellets Figure 2 The spatial grid is divided in the manner shown, with 5 radial nodes and 100 axial nodes.

[0085] According to the fuel cycle management parameters, the ignition consumption program is used to calculate the atomic number N of fission products (including actinides) changing with time. i (t). Using the interface processing program, by reading the database parameters of the ignition consumption program, the reaction cross section parameters, fission yield, decay parameters and other data are obtained, and the fission product production rate B of nuclide i that changes with time is given. i(t). Using I-131, Xe-133 and Xe-135 as target nuclides, Figure 3 The curve showing the change of the number of nuclide atoms in the fuel rod pellet over time is shown. Figure 4 The graph shows the variation of the production rate of nuclides in the fuel rod pellets over time.

[0086] The spatial distribution of power and burnup in the fuel rods is calculated based on the core axial power distribution and the radial power distribution after considering the fuel rod self-shielding effect (zero-order modified Bessel function).

[0087] Power distribution: Fuel consumption distribution:

[0088] Next, calculate the temperature distribution of the fuel rods. Considering the large aspect ratio of the fuel rods, ignoring the axial heat transfer and only considering the radial heat transfer, the single-channel model is used to calculate the temperature distribution of the fuel rods. The calculation results are as follows: Figure 5 shown.

[0089] Next, the fission product release ratio R / B from the fuel pellet to the cladding gap is calculated. The nuclide diffusion coefficient D within the grid is calculated based on the spatial power and temperature distributions. The diffusion coefficients of different nuclides are determined using the relationship D(Xe) = D(Kr) = D(I) = D(Br) / 20 = D(Te) / 4 = 2D(Cs). Based on the spatial burnup and temperature distributions, the S / V ratio within the grid is calculated. The calculation process considers the impact of the combined bubble effect on the increased fission product release:

[0090] When the grid temperature > T link When bubbles coalesce, S / V=650cm -1 , otherwise S / V=120 -1 .

[0091] Taking into account the difference in half-life of nuclides, the calculation model of radionuclide release is expressed as:

[0092]

[0093] Calculate the release yield ratio of a given nuclide i at constant temperature and burnup for all nodes in the fuel rod (5 radial nodes, 100 axial nodes):

[0094]

[0095] The equilibrium relationship of the accumulation of fission products in the gap between the fuel cladding is expressed by the differential equation:

[0096]

[0097] The accumulation curves of I-131, Xe-133 and Xe-135 in the cladding gap over time are as follows: Figure 6 shown.

[0098] By comparing the calculated nuclide inventory data within the fuel cladding with the nuclide inventory within the fuel pellets, the release fraction of each nuclide within the interstitial space of the fuel cladding can be determined. Taking the Xe isotope in this example, the calculated release fractions for Xe-131m are 0.2%, Xe-133 are 0.4%, Xe-133m are 0.8%, Xe-135 are 2.6%, Xe-135m are 2.0%, Xe-137 are 0.6%, and Xe-138 are 1.0%.

[0099] It can be seen that the method provided in the embodiment can give specific release fractions of different nuclides. Compared with the traditional method of assuming that the Xe isotope release fraction is 5%, it is closer to engineering practice and can provide a more accurate and reasonable reference basis for fuel management, fuel assembly design, safety design and management, and accident consequence assessment.

[0100] The purpose of the above embodiments is to provide a further detailed description of the present invention in conjunction with the accompanying drawings so that those skilled in the art can understand the technical concept of the present invention. Within the scope of the present invention, optimization or equivalent replacement of the method steps involved, as well as combination of implementation methods in different embodiments without conflict of principle, all fall within the scope of protection of the present invention.

Claims

1. A method for calculating the fission product accumulation in the fuel cladding gap, characterized in that: The following steps are involved: a) Provide the fuel cycle management parameters, fuel assembly design parameters and primary system design parameters of the target reactor; b) providing an ignition consumption calculation model, calculating the core accumulation of the target nuclide i over time based on the parameters provided in step a), and further calculating the generation rate B of the target nuclide i , where B is the production rate; for the production rate Bi of the target nuclide i, , where Ni(t) is the atomic number, is the rate at which the target nuclide i is produced by fission of the fissionable nuclide j, is the rate at which the target nuclide i is produced by the activation reaction of nuclide k with neutrons, is the production rate of the target nuclide i by the radioactive decay of the nuclide m, is the radiation capture disappearance rate of the target nuclide i; c) Calculate the power distribution p(r,z) and burnup distribution Bu(r,z,t) within the fuel rod, where p is the power, Bu is the burnup distribution, r is the radial position of the calculation node, z is the axial position of the calculation node, and t is time; d) Calculate the radial and axial temperature distribution inside the fuel rod based on p(r,z); e) Calculate the release ratio (R / B) of the target nuclide from the pellet to the cladding gap by combining the burnup distribution and the temperature distribution. i , where R is the number of atoms released in a given space per unit time; f) Binding of different target nuclides (R / B) i The equilibrium relationship of the fission product accumulation in the fuel-cladding gap is calculated to obtain the fission product accumulation in the fuel-cladding gap over time under the operating condition. The equilibrium relationship of the fission product accumulation in the fuel-cladding gap is calculated using the following differential equation: ， in, ; Where P is power, V is volume, rod is the fuel rod as a whole, node is the axial node, ring is the radial ring, and A is the g,i (t) is the radioactive accumulation of nuclide i, A g,i (t) is the radioactive inventory of parent nucleus j of nuclide i in the fuel cladding.

2. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 1, characterized in that: In step a), the fuel cycle management parameters include fuel enrichment, power density, irradiation cycle length, and core axial power distribution; the fuel assembly design parameters include material composition and fuel rod size parameters; and the primary loop system design parameters include pressure, temperature, and flow rate.

3. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 1, characterized in that: The ignition consumption calculation model includes an ignition consumption parameter database, and the ignition consumption parameter database includes reaction cross section parameters, fission yields, and decay parameters.

4. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 1, characterized in that: In the step c), the power distribution includes radial power distribution and axial power distribution, wherein: The radial power distribution , where g(r) is the radial power distribution function of the fuel rod, C is the normalization constant, and I0(x) is the first kind of zero-order modified Bessel function; , where ε is the fuel enrichment percentage, ρ U is the U density in the fuel rod, R fuel is the fuel pellet radius; The axial power distribution f(z) adopts the axial power distribution of the fuel rods in the reactor design, or adopts , where L is the length of the fuel rod; , where P rod is the average power of the fuel rod; ,in is the fuel volume at the (r,z) node.

5. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 1, characterized in that: In step d), the temperature distribution of the fuel rods is calculated using the following model: ， ， ， ， ； Among them, T b is the mainstream temperature of the single-channel coolant, T in is the inlet temperature of a single channel, T cs is the outer surface temperature of the cladding, T ci is the inner surface temperature of the cladding, T fs is the fuel pellet surface temperature, T p is the temperature distribution inside the fuel pellet, D0 is the equivalent diameter of a single channel, A f is the cross-sectional area of ​​a single channel, q n is the radial heat flux of the fuel rod, G is the mass flux of the coolant in a single channel, C p is the constant pressure heat capacity of the coolant, h f is the convective heat transfer coefficient, q i is the radial linear power density of the fuel rod, r o is the outer diameter of the fuel rod cladding, r i is the inner diameter of the fuel rod cladding, k c is the thermal conductivity of the fuel rod cladding, k gas is the thermal conductivity of the gas in the cladding gap, q v is the volume heat release rate of the fuel rod, k fuel is the thermal conductivity of the fuel pellet.

6. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 1, characterized in that: In step e), R / B is calculated by solving the equation We can obtain , where C is the nuclide concentration in the fuel, λ is the nuclide decay constant, and D is the nuclide diffusion coefficient.

7. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 1 or 6, characterized in that: For the target nuclide i, ； in, is the release production ratio of nuclide i at the node position (r, z); is the ratio of the surface area to the volume at the node position (r, z); F i is the conversion factor for long-lived nuclides, ;H j,i is the correction factor for the diffusion ability of parent nucleus j to nuclide i; , where D is the diffusion coefficient of the nuclide and a is the radius of the ideal sphere equivalent to the fuel grain.

8. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 7, characterized in that: D = D1(T) + D2(T,F) + D3(F), where D1 is the intrinsic diffusion coefficient, D2 is the diffusion coefficient of the vacancy mechanism increased by irradiation, D3 is the non-thermal diffusion, T is the fuel pellet temperature, and F is the fission rate.

9. The method for calculating the fission product accumulation in the fuel cladding gap according to claim 7, characterized in that: There is a sudden change in temperature T caused by the combined effect of bubbles link , in the calculation, the temperature is greater than T link and less than T link hour Different values ​​are used respectively, among which: 。 10. A computing device comprising a memory and a processor, characterized in that: The memory stores a calculation program, and when the calculation program is executed by the processor, the method for calculating the amount of fission product accumulation in the gap between fuel claddings as claimed in any one of claims 1 to 9 can be implemented.

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