A design and analysis method for a small pressurized water reactor core based on accident-tolerant fuel

Abstract

The invention discloses a small pressurized water reactor core design and analysis method based on accident fault-tolerant fuel, comprising the following steps: constructing a sea-based small pressurized water reactor core component of a fault-tolerant fuel system, constructing a gate element and component calculation model, performing a small group parameterization process on the gate element and component calculation model, analyzing the steady-state characteristics of the sea-based small pressurized water reactor and analyzing the transient characteristics of the sea-based small pressurized water reactor; the invention sets a sea-based small pressurized water reactor with accident fault-tolerant fuel, analyzes its residual reactivity, reactivity coefficient, power distribution and controlled poison value in a steady state, and simultaneously analyzes the change law of the reactivity, power level, moderator temperature density and Doppler temperature of four reactor cores under the rod-bombing condition in a transient accident with time, provides a reference for the subsequent ATF sea-based small pressurized water reactor core design, facilitates people to select fuel according to different needs, and has important significance and engineering practical value.

Description

Technical Field

[0001] The invention relates to the technical field of small pressurized water reactor core design, and in particular to a small pressurized water reactor core design and analysis method based on accident fault-tolerant fuel. Background Art

[0002] Since the 1990s, Small Modular Reactor (SMR) has attracted widespread attention due to its inherent safety, low investment and flexibility. As a small, movable floating reactor used to develop the marine economy, sea-based small pressurized water reactor has become a key research object in various countries. It can provide electricity, heat and fresh water resources for offshore natural gas and oil production platforms, and can also supply efficient and safe energy for remote islands. At the same time, it is of great significance in safeguarding national maritime rights and interests.

[0003] Currently, most pressurized water reactors in the world use uranium dioxide fuel and zirconium alloy cladding. Although this type of fuel system is the most widely used, it still has some defects. For example, when the reactor is running, the fuel center temperature is too high due to the poor thermal conductivity of the fuel, which can easily cause a meltdown accident. The hydrogen explosion that may be caused by the zirconium-water reaction is also an issue that needs to be considered during reactor safety design. Therefore, accident tolerant fuel (ATF) was proposed based on this concept.

[0004] At present, accident-tolerant fuel can be divided into accident-tolerant fuel core blocks and accident-tolerant cladding materials. Accident-tolerant fuel is a new type of fuel system developed to improve the ability of fuel elements to resist major accidents. Compared with traditional nuclear fuel, the new fuel system can resist serious accidents for a longer period of time. At the same time, it can maintain or improve its performance under normal operating conditions. Due to its material properties, it can slow down the progress of accidents and buy more precious time for people to take emergency measures, thereby greatly reducing the risk of radioactive substances breaking through safety barriers and leaking into the environment. It has important significance and engineering practical value.

[0005] However, most of the existing small pressurized water reactor core design and analysis methods have a single process, and usually conduct thermal hydraulic analysis based on the performance of a single accident-tolerant fuel under an accident. The relevant research on sea-based small pressurized water reactors loaded with accident-tolerant fuel is not perfect enough, resulting in the inability to derive the impact of different ATF fuel systems on the neutronic performance of small pressurized water reactors and the improvement of safety performance. Therefore, the present invention proposes a small pressurized water reactor core design and analysis method based on accident-tolerant fuel to solve the problems existing in the prior art. Summary of the invention

[0006] In view of the above problems, the purpose of the present invention is to propose a small pressurized water reactor core design and analysis method based on accident-tolerant fuel, so as to solve the problem that the existing small pressurized water reactor core design and analysis methods are unable to derive the impact of different ATF fuel systems on the neutronic performance of the small pressurized water reactor and the improvement of safety performance.

[0007] In order to achieve the purpose of the present invention, the present invention is implemented by the following technical scheme: a small pressurized water reactor core design and analysis method based on accident fault-tolerant fuel, comprising the following steps:

[0008] Step 1: First, analyze the advantages and characteristics of different fuel pellet materials, and select the fuel pellet material with the greatest potential as the spare pellet of the accident-tolerant fuel. Then, analyze the accident-tolerant capabilities of different cladding materials, and select the cladding material with the strongest accident-tolerant capability as the spare fuel cladding. The selected pellets and fuel cladding are combined into a sea-based small pressurized water reactor core assembly of the fault-tolerant fuel system.

[0009] Step 2: First, determine the fuel grid element, control rod grid element and burnable poison grid element according to the combined sea-based small pressurized water reactor core assembly, then use the fuel grid element and burnable poison grid element to build a burnable poison assembly, and use the fuel grid element and control rod grid element to build a control assembly, then build a reflector assembly according to the burnable poison assembly and the control assembly, and then build a grid element and assembly calculation model based on the ACPR50S calculation model;

[0010] Step 3: First, the grid element and component calculation model are modeled and calculated through the two-dimensional component program HELIOS to obtain the group constant, and then the component program output file is read by GenPMAXS, and the PMAXS file of the macroscopic cross-section structure is generated after processing for PARCS to read, and then the real flux of the sea-based small pressurized water reactor core is calculated, and then the super grid element model is established for the reflector component, and the effective uniform reflector ADF is obtained to achieve the homogenization of the grid element and component calculation model;

[0011] Step 4: Based on the calculation model of gate elements and components, first analyze the influence of factors such as fuel temperature, moderator temperature, moderator density, coolant cavitation coefficient and power coefficient on reactivity, and then evaluate the critical safety characteristics of the sea-based small pressurized water reactor based on accident-tolerant fuel under steady-state conditions through the analysis method of nuclear-thermal coupling according to the reactivity feedback coefficient, residual reactivity and active area power flattening parameters, so as to analyze the steady-state characteristics of the sea-based small pressurized water reactor core components of the fault-tolerant fuel system;

[0012] Step 5: First, based on the calculation model of gate elements and components and on the basis of steady-state characteristics analysis, a reactivity introduction accident is designed. The transient neutronics and thermal-hydraulic characteristics of the sea-based small pressurized water reactor under typical accident conditions are analyzed and evaluated from the core reactivity, Doppler radial temperature distribution, moderator temperature and power flattening conditions, so as to realize the transient safety characteristics analysis of the sea-based small pressurized water reactor core components of the fault-tolerant fuel system.

[0013] A further improvement is that in step 1, the analysis of the characteristics of different fuel pellet materials is specifically an analysis of the density, melting point and thermal conductivity of heavy metals.

[0014] A further improvement is that in step three, the PMAXS file represents the macroscopic cross section as a function of state variables, wherein the state variables include control rods, coolant density, coolant soluble poison concentration, fuel temperature, coolant temperature, coolant impurities, moderator density, moderator soluble poison concentration, moderator temperature and moderator impurities.

[0015] Further improvement lies in: in the step three, the specific steps to obtain the effective uniform reflection layer ADF are: first establish the reflection layer super grid element model through HELIOS, solve the one-dimensional diffusion equation in the reflection layer area to determine the surface average flux of the uniform reflection layer, the neutron flow and flux in the east and west directions, and then obtain the effective uniform reflection layer ADF through the reflection layer ADF 1D calculation module of GenPMAXS.

[0016] A further improvement is that in step 4, the fuel temperature reactivity coefficient refers to the change in core reactivity caused by a change of 1K in fuel temperature.

[0017] A further improvement is that in step 4, the fuel temperature range is selected to be 900-1800K, and the core k at the corresponding temperature is calculated every 100K. eff .

[0018] Further improvement is that: in step 4, the moderator temperature coefficient refers to the change in core reactivity caused by a 1K change in the moderator, the moderator temperature range is selected to be 520-610K, and the core k under different moderator temperatures and their corresponding densities is calculated every 10K eff .

[0019] A further improvement is that in step 4, the coolant cavitation coefficient refers to the volume fraction of cavitation contained in the coolant, the cavitation fraction is 0%-100%, and the core k of the core under different cavitation fractions is calculated every 10%. eff .

[0020] The beneficial effects of the present invention are as follows: the present invention designs a sea-based small pressurized water reactor with BeO+UO2-SiC accident-tolerant fuel, analyzes its residual reactivity, reactivity coefficient, power distribution and controlled poison value in a steady state, and simultaneously analyzes the variation of the reactivity, power level, moderator temperature density and Doppler temperature of four reactor cores under the rod-bombing condition in a transient accident with time, so as to provide a reference for the subsequent ATF sea-based small pressurized water reactor core design, and through the coupling analysis between neutronics and thermal hydraulics, carries out the characteristic study of the sea-based small pressurized water reactor based on different accident-tolerant fuels under steady-state and transient conditions, analyzes its safety and economy, and facilitates people to select fuel according to different needs, which has important significance and engineering practical value. BRIEF DESCRIPTION OF THE DRAWINGS

[0021] Figure 1 It is a schematic flow chart of the method of the present invention;

[0022] Figure 2 is a schematic diagram of the geometry of the fuel grid element of the present invention;

[0023] Figure 3 is a geometrical schematic diagram of a control rod grid element of the present invention;

[0024] Figure 4 It is a geometrical schematic diagram of the combustible poison cell of the present invention;

[0025] Figure 5 It is a schematic diagram of the geometric arrangement of the burnable poison assembly of the present invention;

[0026] Figure 6 is a schematic diagram of the geometric arrangement of the control components of the present invention;

[0027] Figure 7 Schematic diagram of the classification of the four types of reflective layers of the present invention;

[0028] Figure 8 It is a schematic diagram of a 1 / 4 core of the ejection rod position information of the present invention. DETAILED DESCRIPTION

[0029] In order to deepen the understanding of the present invention, the present invention will be further described in detail below in conjunction with examples. The examples are only used to explain the present invention and do not constitute a limitation on the protection scope of the present invention.

[0030] See also Figure 1 , Figure 2 , Figure 3 , Figure 4 , Figure 5 , Figure 6 , Figure 7 , Figure 8 This embodiment provides a small pressurized water reactor core design and analysis method based on accident-tolerant fuel, comprising the following steps:

[0031] Step 1: First, analyze the advantages and characteristics of different fuel pellet materials. The characteristic analysis of different fuel pellet materials is specifically the analysis of heavy metal density, melting point and thermal conductivity, and select the fuel pellet material with the most potential as the spare pellet of accident fault-tolerant fuel. In this embodiment, uranium dioxide-based accident fault-tolerant fuel UO2+BeO pellets are selected. There are two main technical routes for upgrading uranium dioxide-based fuel. On the one hand, high thermal conductivity materials are added to enhance the thermal conductivity of UO2, and on the other hand, large-grained UO2 is prepared to reduce heat conduction loss. Then, the accident fault tolerance of different cladding materials is analyzed, and the cladding material with the strongest accident fault tolerance is selected as the spare fuel cladding. In this embodiment, SiC composite material is selected as the fuel cladding, and the selected pellets and fuel cladding are combined into a sea-based small pressurized water reactor core component UO2+BeO-SiC of the fault-tolerant fuel system;

[0032] Step 2: First, determine the fuel grid element, control rod grid element and burnable poison grid element according to the assembled sea-based small pressurized water reactor core assembly, where the fuel grid element has a size of 1.25984cm×1.25984cm, and the geometric diagram is as follows: Figure 2 As shown in the figure, the fuel area is divided into 9 circles, with radii of 0.1cm, 0.15cm, 0.2cm, 0.25cm, 0.3cm, 0.35cm, 0.38cm, 0.4cm, and 0.424575cm respectively. The thickness of the helium gap is 82.55, the inner diameter of the cladding is 0.43283cm, and the outer diameter is 0.48998cm. The light water part is divided into 5 circles, with radii of 0.495cm, 0.50cm, 0.53cm, 0.56cm, and 0.60cm respectively.

[0033] The size of the control rod grid element is 1.25984cm×1.25984cm, and the geometric diagram is as follows Figure 3 As shown, the control rod Ag-In-Cd in the figure is divided into 9 circles, with radii of 0.05cm, 0.1cm, 0.2cm, 0.25cm, 0.3cm, 0.35cm, 0.48cm, 0.52cm, and 0.571cm respectively. When the rod is not inserted, it is filled with light water as a moderator. The inner diameter of the cladding (SiC) is 0.571cm and the outer diameter is 0.6130cm. The light water at the four corners of the grid element in the figure is refined respectively, and the four sides are evenly divided into 8 segments according to the coordinates, and the length of each segment is 0.15748cm.

[0034] The size of the burnable poison (WABA) grid element is 1.25984cm×1.25984cm. The density of the burnable poison Al2O3-B4C is 3.1. The geometric diagram is as follows: Figure 4As shown, 7 areas are marked in the figure to represent the composition of the grid element. Area 1 is light water, divided into two circles, with radii of 0.1 cm and 0.286 cm respectively. Area 2 is Zr-4 alloy cladding with a radius of 0.353 cm. Area 3 is Al2O3-B4C with a radius of 0.404 cm. Area 4 is Zr-4 alloy cladding with a radius of 0.484 cm. Area 5 is light water with a radius of 0.596 cm. Area 6 is Zr-4 alloy cladding with a radius of 0.615 cm. Area 7 is light water with a radius of 0.62992 cm. The outermost moderator part is the same as the moderator refinement method in the four corners of the control rod grid element.

[0035] The fuel grid element and the burnable poison grid element are then used to construct a burnable poison (WABA) assembly, the size of which is 21.41728cm×21.41728cm. Figure 5 As shown, a control assembly is constructed using a fuel grid element and a control rod grid element, and the assembly size is 21.41728cm×21.41728cm, as shown in FIG. Figure 6 As shown, the reflector assembly is then constructed based on the burnable poison assembly and the control assembly. Considering the difference between the reflector and the adjacent assembly materials, the distance between the reflector and the active area, etc., taking a quarter of the core as an example, the reflector assembly is divided into the following four categories: Figure 7 As shown, the gate cell and component calculation models are then constructed based on the ACPR50S calculation model;

[0036] Step 3: First, the gate element and component calculation model are modeled and calculated through the two-dimensional component program HELIOS to obtain the group constant, and then the component program output file is read by GenPMAXS. After processing, the PMAXS file of the macroscopic cross-section structure is generated for PARCS to read. The PMAXS file represents the macroscopic cross-section as a function of state variables. The state variables usually include control rods, coolant density, coolant soluble poison concentration, fuel temperature, coolant temperature, coolant impurities, moderator density, moderator soluble poison concentration, moderator temperature, and moderator impurities;

[0037] The macroscopic cross section is also a function of burnup and other historical variables. Since the macroscopic cross section strongly depends on the historical effect, PMAXS also introduces control rod history (HCR), coolant density history (HDC), coolant soluble poison history (HPC), fuel temperature history (HTF), and coolant temperature history (HTC). The historical variables and burnup jointly determine the historical state.

[0038] In order to reflect the real flux distribution in the core, the generalized homogenization theory proposed by KK Smith is adopted, and the component homogenization cross section (XS) and discontinuity factor (ADF) are applied to the nodal method to ensure the conservation of the neutron reaction rate of each nodal block and the interface flow between nodal blocks. s,gis defined as the ratio of the non-uniform flux to the uniform flux,

[0039]

[0040] In the formula, represents the non-uniform flux at the node interface, represents uniform flux, g is the number of energy groups, s is the node interface, and the non-uniform flux is directly calculated by the two-dimensional component program HELIOS. Since the fuel assembly is a total reflection boundary condition, the neutron flux before and after homogenization is flat and constant, so the average flux of the assembly is equal to the neutron flux on the surface of the homogenized assembly. Under the total reflection boundary condition, ADF can be calculated using the following formula:

[0041]

[0042] in, Uniform flux for the components.

[0043] For the reflector components, a super-grid element model is established. First, the reflector super-grid element model is established through HELIOS, and the one-dimensional diffusion equation in the reflector area is solved to determine the surface average flux of the uniform reflector, the neutron flow and flux in the east and west directions, and then the reflector ADF 1D calculation module of GenPMAXS is used to obtain the effective uniform reflector ADF, so as to achieve the homogenization of the grid element and component calculation model;

[0044] Step 4: Based on the calculation model of gate elements and components, first analyze the influence of factors such as fuel temperature, moderator temperature, moderator density, coolant cavitation coefficient and power coefficient on reactivity, and then evaluate the critical safety characteristics of the sea-based small pressurized water reactor based on accident-tolerant fuel under steady-state conditions through the analysis method of nuclear-thermal coupling according to the reactivity feedback coefficient, residual reactivity and active zone power flattening parameters, and analyze the steady-state characteristics of the sea-based small pressurized water reactor core components of the fault-tolerant fuel system. The abundance of the burnable poison (WABA) component B-10 is selected to be 5%;

[0045] The fuel temperature reactivity coefficient refers to the change in core reactivity caused by a 1K change in fuel temperature. As the fuel temperature rises, resonance broadening occurs due to the Doppler effect, and the absorption cross section of the nuclide U-238 with resonance characteristics in the fuel decreases, the resonance peak decreases, and the energy self-screening effect weakens. At this time, the number of neutrons in the resonance peak increases, and the probability of neutrons escaping the resonance absorption decreases, which in turn makes the core k eff Decreases, the reactivity decreases, that is:

[0046]

[0047] in, represents the fuel temperature reactivity coefficient, P represents the core reactivity, and T represents the fuel temperature in the core;

[0048] Since the core has a high internal temperature during operation, the fuel temperature range is selected to be 900-1800K, and the k at the corresponding temperature is calculated every 100K. eff ,When calculating the fuel temperature coefficient, only the effect of fuel temperature on core reactivity is considered, and the effect of rising fuel temperature on coolant is ignored;

[0049] The fuel temperature reactivity coefficient (pcm / K) is -2.74530

[0050] The moderator temperature coefficient refers to the change in core reactivity caused by a 1K change in moderator temperature;

[0051]

[0052] in, represents the moderator temperature coefficient, P represents the core reactivity, and T represents the moderator temperature in the core;

[0053] Since the pressure of a PWR is usually maintained at about 150 atmospheres, water can be heated to about 616K without boiling under this pressure. Considering that the temperature variation of the moderator in a small reactor is small, the moderator temperature range is selected to be 520-610K, and the calculation is performed every 10K. Since the moderator temperature and density change synchronously, the core k under different moderator temperatures and their corresponding densities is calculated every 10K. eff ;

[0054] The temperature reactivity coefficient of the moderator (pcm / K) is -77.09

[0055] The cavitation fraction refers to the volume fraction of cavitation contained in the coolant (%). When the coolant of a sea-based small reactor is lost, the change in its reactivity has a significant impact on the safe operation of the reactor. Similar to the reactors operating on land, the cavitation value of a sea-based reactor refers to the change in reactivity caused by a 1% change in the volume fraction of cavitation in the coolant.

[0056]

[0057] Among them, it means is the cavitation coefficient, x represents the cavitation volume fraction;

[0058] Investigate the effect of different coolant cavitation fractions on reactivity. The cavitation fraction is 0%-100%, and the calculation is performed every 10%. The core k of the reactor core under different cavitation fractions is calculated. eff , giving a cavitation coefficient (pcm / %) of -331.100.

[0059] In safe operation and analysis, the power reactivity coefficient, referred to as the power coefficient, is more practical. The power coefficient refers to the change in core reactivity caused by a 1% change in the reactor power level. It is a comprehensive effect of reactivity feedback including fuel temperature coefficient, moderator temperature coefficient, cavitation coefficient, etc. No matter which reactivity coefficient changes, it will eventually affect the core reactivity through power changes. As an overall coefficient, as long as the power coefficient is negative, the sea-based small reactor can be considered safe.

[0060]

[0061] Where a P represents the power factor, is the fuel temperature coefficient, represents the moderator temperature coefficient, represents the cavitation coefficient;

[0062] Under the rated power of 300MWth, the power density of the core active area is calculated when the control rods are fully raised, and the radial normalized power distribution of the core power (P assembly / P average) and the radial power peak factor are given.

[0063] Step 5: First, based on the calculation model of gate elements and components and on the basis of steady-state characteristic analysis, a reactivity introduction accident is designed, and transient neutronics and thermal hydraulics characteristics of the sea-based small pressurized water reactor under typical accident conditions are evaluated from the core reactivity, Doppler radial temperature distribution, moderator temperature and power flattening conditions, so as to realize the transient safety characteristics analysis of the sea-based small pressurized water reactor core components of the fault-tolerant fuel system;

[0064] The radial schematic diagram of the control rod assembly is shown in Figure 8 As shown, when the control rod assembly is in a fully inserted state, the edge of the bottom end of the control rod is 10 cm away from the bottom of the core. For a fully withdrawn control rod assembly, the bottom end of the control rod absorber is 260 cm away from the bottom of the core. In units of step length, the fully inserted state corresponds to 0 step length, and the fully withdrawn state corresponds to 260 step lengths, and the length of each step length is 1 cm. The embodiment simulates two groups of rod ejection accidents, namely, group A with the smallest control value and group B with the largest control value in the hot full power state:

[0065] (Casea), Group A control rods are bounced from the initial 0 step length to 100 step length, a total of 100 cm;

[0066] (Caseb), the length of group B control rods increases from the initial 100 steps to 260 steps when completely ejected from the core, a total of 160 cm.

[0067] The design rod ejection time is 0.1s, the transient duration of nuclear thermal coupling calculation is 10s, the initial time step is calculated every 0.01s, and the new time step is switched to 0.02s after 1s of calculation. A total of 551 calculation points are generated in 10s;

[0068] In the Case a case, the core power level reaches the maximum power at the transient time t=0.19s, which can reach 1.64 times the rated power; the overall change of the moderator temperature increases from 320.2℃ to a maximum of 322.62℃; the overall change of the Doppler temperature increases from 479.35℃ to a maximum of 516℃.

[0069] In the Case b case, the core power level reaches the maximum power at the transient time t=0.08s, which can reach 2.31 times the rated power; the overall change of the moderator temperature increases from 320.2℃ to a maximum of 322.62℃; the overall change of the Doppler temperature increases from 479.35℃ to a maximum of 516℃.

[0070] The above shows and describes the basic principles, main features and advantages of the present invention. It should be understood by those skilled in the art that the present invention is not limited to the above embodiments. The above embodiments and descriptions are only for explaining the principles of the present invention. Without departing from the spirit and scope of the present invention, the present invention may have various changes and improvements, which fall within the scope of the present invention. The scope of protection of the present invention is defined by the attached claims and their equivalents.

Claims

1. A design and analysis method for a small pressurized water reactor core based on accident-tolerant fuel, characterized in that: The following steps are involved: Step 1: First, analyze the advantages and characteristics of different fuel pellet materials, and select the fuel pellet material with the greatest potential as the spare pellet of the accident-tolerant fuel. Then, analyze the accident-tolerant capabilities of different cladding materials, and select the cladding material with the strongest accident-tolerant capability as the spare fuel cladding. The selected pellets and fuel cladding are combined into a sea-based small pressurized water reactor core assembly of the fault-tolerant fuel system. Step 2: First, determine the fuel grid element, control rod grid element and burnable poison grid element according to the combined sea-based small pressurized water reactor core assembly, then use the fuel grid element and burnable poison grid element to build a burnable poison assembly, and use the fuel grid element and control rod grid element to build a control assembly, then build a reflector assembly according to the burnable poison assembly and the control assembly, and then build a grid element and assembly calculation model based on the ACPR50S calculation model; Step 3: First, the grid element and component calculation model are modeled and calculated through the two-dimensional component program HELIOS to obtain the group constant, and then the component program output file is read by GenPMAXS, and the PMAXS file of the macroscopic cross-section structure is generated after processing for PARCS to read, and then the real flux of the sea-based small pressurized water reactor core is calculated, and then the super grid element model is established for the reflector component, and the effective uniform reflector ADF is obtained to achieve the homogenization of the grid element and component calculation model; Step 4: Based on the calculation model of gate elements and components, first analyze the influence of factors such as fuel temperature, moderator temperature, moderator density, coolant cavitation coefficient and power coefficient on reactivity, and then evaluate the critical safety characteristics of the sea-based small pressurized water reactor based on accident-tolerant fuel under steady-state conditions through the analysis method of nuclear-thermal coupling according to the reactivity feedback coefficient, residual reactivity and active area power flattening parameters, so as to analyze the steady-state characteristics of the sea-based small pressurized water reactor core components of the fault-tolerant fuel system; Step 5: First, based on the calculation model of gate elements and components and on the basis of steady-state characteristics analysis, a reactivity introduction accident is designed. The transient neutronics and thermal-hydraulic characteristics of the sea-based small pressurized water reactor under typical accident conditions are analyzed and evaluated from the core reactivity, Doppler radial temperature distribution, moderator temperature and power flattening conditions, so as to realize the transient safety characteristics analysis of the sea-based small pressurized water reactor core components of the fault-tolerant fuel system.

2. The method for designing and analyzing a small pressurized water reactor core based on accident-tolerant fuel according to claim 1, characterized in that: In the step 1, the analysis of the characteristics of different fuel pellet materials is specifically the analysis of the density, melting point and thermal conductivity of heavy metals.

3. The method for designing and analyzing a small pressurized water reactor core based on accident-tolerant fuel according to claim 1, characterized in that: In step three, the PMAXS file represents the macroscopic cross section as a function of state variables, wherein the state variables include control rods, coolant density, coolant soluble poison concentration, fuel temperature, coolant temperature, coolant impurities, moderator density, moderator soluble poison concentration, moderator temperature, and moderator impurities.

4. The method for designing and analyzing a small pressurized water reactor core based on accident-tolerant fuel according to claim 1, characterized in that: In step three, the specific steps for obtaining the effective uniform reflection layer ADF are: firstly establish the reflection layer super grid element model through HELIOS, solve the one-dimensional diffusion equation in the reflection layer area to determine the surface average flux of the uniform reflection layer, the neutron flux and flux in the east and west directions, and then obtain the effective uniform reflection layer ADF through the reflection layer ADF 1D calculation module of GenPMAXS.

5. The method for designing and analyzing a small pressurized water reactor core based on accident-tolerant fuel according to claim 1, characterized in that: In step 4, the fuel temperature reactivity coefficient refers to the change in core reactivity caused by a change of 1K in fuel temperature.

6. The method for designing and analyzing a small pressurized water reactor core based on accident-tolerant fuel according to claim 1, characterized in that: In step 4, the fuel temperature range is selected to be 900-1800K, and the core k at the corresponding temperature is calculated every 100K. eff .

7. The method for designing and analyzing a small pressurized water reactor core based on accident-tolerant fuel according to claim 1, characterized in that: In step 4, the moderator temperature coefficient refers to the change in core reactivity caused by a 1K change in the moderator. The moderator temperature range is 520-610K, and the core k under different moderator temperatures and corresponding densities is calculated every 10K. eff .

8. The method for designing and analyzing a small pressurized water reactor core based on accident-tolerant fuel according to claim 1, characterized in that: In step 4, the coolant cavitation coefficient refers to the volume fraction of cavitation contained in the coolant, the cavitation fraction is 0%-100%, and the core k of the core under different cavitation fractions is calculated every 10%. eff .

Images