CN-121997820-A - Core physical optimization calculation method under nuclear power unit partial ring operation condition

CN121997820ACN 121997820 ACN121997820 ACN 121997820ACN-121997820-A

Abstract

The invention relates to the field of reactor core physical computation, in particular to a reactor core physical optimization computation method under the operating condition of a nuclear power unit deflection loop, which comprises the following steps of constructing a geometric model and a reactor core grid model of a reactor in fluid dynamics software; the method comprises the steps of establishing a porous medium model in fluid dynamics software, setting boundary conditions in the fluid dynamics software and carrying out thermodynamic and hydraulic calculation, establishing a thermodynamic and physical coupling interface, carrying out physical calculation in physical calculation software by taking a thermodynamic calculation result of the fluid dynamics software as input, carrying out thermodynamic and hydraulic calculation by taking three-dimensional power distribution of the physical calculation software as input of the fluid dynamics software to obtain an optimized power peak factor on a fuel assembly and a fuel assembly volumetric power peak factor, and repeating the steps three to five until the power peak factor converges. According to the invention, the reactor core physical parameters under the operating condition of the deflection ring are calculated, so that the control capability of the safe operation of the reactor is further improved.

Inventors

LI YAO
DONG WENCHANG
PAN DENG
PENG YIPENG
LUO DENGYU
LV ZHUMEI
Hua Xinchao

Assignees

核动力运行研究所

Dates

Publication Date: 20260508
Application Date: 20260115

Claims (8)

1. A core physical optimization calculation method under a nuclear power unit partial ring operation condition is characterized by comprising the following steps: Step one, constructing a geometrical model and a reactor core grid model of a reactor in fluid dynamics software; step two, constructing a porous medium model in fluid dynamics software; Setting boundary conditions in fluid dynamics software, carrying out thermodynamic and hydraulic calculation, and carrying out normalization processing on the flow of each fuel assembly of the reactor core through average flow on the flow of the reactor core inlet to obtain reactor core inlet flow distribution; Step four, constructing a thermodynamic and physical coupling interface, taking a thermodynamic calculation result of fluid dynamics software as input, and carrying out physical calculation in the physical calculation software to obtain a reactivity coefficient of a reactor core and three-dimensional power distribution, wherein the three-dimensional power distribution comprises a power peak factor Kq on a fuel assembly and a fuel assembly volume power peak factor Kv; Taking three-dimensional power distribution of physical calculation software as fluid dynamics software input, carrying out thermodynamic and hydraulic calculation again to obtain updated calculated reactor core inlet flow distribution, taking the result as input, and carrying out physical calculation on the physical calculation software again to obtain optimally calculated fuel assembly upper power peak factor Kq and fuel assembly volume power peak factor Kv; and step six, comparing the power peak factor Kq on the fuel assembly and the power peak factor Kv on the fuel assembly in the step four and the step five, and repeating the steps three to five until the deviation of the power peak factor Kq on the fuel assembly and the power peak factor Kv on the fuel assembly meets the precision requirement.
2. The method for optimizing and calculating the physical reactor core under the operating condition of the eccentric ring of the nuclear power unit according to claim 1 is characterized in that the geometric model of the reactor in the first step comprises a core fuel assembly, a core coaming, a core basket, a fuel assembly supporting seat, a flow isolation ring, a protection pipe assembly, the inner wall surface of a reactor pressure vessel and a coolant flow inlet and outlet area, and after the geometric model is established, the geometric model is meshed to construct a core mesh model.
3. The core physical optimization calculation method under the nuclear power unit deflection ring operation condition according to claim 2 is characterized in that when a geometric model is built in the first step, the pressure vessel model is adjusted or simplified on the premise that the integral flow characteristic in the pressure vessel is not affected, the ellipsoidal bottom head of a core basket model is equivalent to 1 opening on the premise that coolant uniformly enters a lower cavity, the large opening and the peripheral small opening are equivalent to 1 opening on the basis that the flow cross section area is equal, and the narrow grooves of the fuel assembly supporting seat model are equivalent to 6 narrow grooves uniformly distributed along the circumferential direction of a cylinder on the premise that the flow area is ensured in the middle of the fuel assembly supporting seat model.
4. The method for optimizing and calculating the physical properties of a reactor core under the operating condition of a deflection ring of a nuclear power unit according to claim 3, wherein the reactor core grid model in the first step adopts unstructured tetrahedral grid type to generate integrated grids, and the grids are encrypted in areas with smaller dimensions such as a lower plate runner gap of a protective tube assembly, wall openings of the lower plate runner gap, narrow grooves of a supporting seat of a fuel assembly and the like and runner turning positions.
5. The method for optimizing and calculating the physical reactor core under the operating condition of the eccentric ring of the nuclear power unit according to claim 4, wherein in the second step, a porous medium model is adopted to simulate the fuel assembly of the reactor core, the cooling water flow area surrounded by the whole reactor core coaming is an independent fluid area, the type of the flow area is set to be a porous medium area, and the porous medium model is required to be provided with porosity, permeability, secondary resistance coefficient and internal heat source thermal power.
6. The method for calculating the physical optimization of the reactor core under the partial ring operation condition of the nuclear power unit according to claim 5 is characterized in that a porous medium model is adopted in the second step to simulate a reactor core fuel assembly, the volume porosity of the porous medium model is equal to the surface porosity, the permeability is defined by Darcy's law and calculated according to actual operation parameters, the flow direction secondary resistance coefficient is set to be 100 times of the radial flow resistance coefficient of the reactor core, the reactor core thermal power is compiled by UDF, the actual operation thermal power of a reactor core area is loaded into an energy transport equation by utilizing the UDF as a transport equation source term, and a reactor core heat release mode is defined to perform simulation calculation of a non-uniform internal heat source.
7. The core physical optimization calculation method under the nuclear power unit partial ring operation condition of claim 6 is characterized in that boundary conditions in the third step are that an inlet of a calculation domain is set to be an inlet connecting pipe of a reactor pressure vessel, an outlet of the calculation domain is set to be an outlet connecting pipe of the reactor pressure vessel, a boundary of fluid is set to be a non-slip heat insulation wall boundary condition on the inner wall surface of the reactor pressure vessel, physical properties such as density, heat conductivity and specific heat capacity of a coolant material change along with temperature, gravity is selected to be considered in the axial direction of the coolant material, the inlet mass flow of the pressure vessel of a failure loop is a negative value, and design data of the inlet mass flow, temperature and outlet pressure of the pressure vessel are converted.
8. The method is characterized in that in the fourth step, a thermotechnical and physical coupling interface is constructed, physical calculation software is KASKAD program, the KASKAD program is modified by using a subroutine for the three-dimensional reactor core neutron calculation aiming at the KASKAD program, the assignment of symmetry parts is modified, the KASKAD program is used for calculating the volume of the core and the subroutine of the unit edge connected with the radial reflecting layer, the interface is written by adopting the FORTRAN language, the flow distribution result obtained by the third step is used for modifying the flow distribution result obtained by the calculation, the relative coolant flow velocity flowing through a fuel assembly is customized in a configuration file, the KASKAD program is used for simulating the subroutine for calculating the fuel consumption, the parameters are correspondingly modified in the debugging, the KASKAD program is modified by using a shutdown failure loop in the main circulating pump according to the actual partial loop operation, the thermal feedback correction coefficient in the configuration file is modified by using the actual operation loop number, the temperature and the flow are modified by using the thermal calculation result of the KASKAD program, and the power compensation bar groups and the adjustment bar positions under different state points are adjusted according to the input power of the actual operation working condition.

Description

Core physical optimization calculation method under nuclear power unit partial ring operation condition Technical Field The invention relates to the field of core physical computation, in particular to a core physical optimization computation method under the operating condition of a nuclear power unit deflection ring. Background And under the normal operation state of the VVER unit, the four loops are operated in parallel. In the event of an abnormality such as a failure of the main pump, the unit is allowed to operate with reduced power in the partial loop operation mode, i.e., the partial loop operation. Under the partial-ring operation condition, on one hand, extremely uneven component flow distribution can cause partial component power and temperature to be higher, so that the operation margin of a unit is reduced, and potential threat is formed to the safety of a reactor core, and on the other hand, the simulation of the existing calculation program to the partial-ring operation condition is rough, so that obvious deviation exists between theoretically calculated power distribution and temperature distribution and measured values, and the real-time monitoring and accurate control of the reactor core can be influenced. Multiple partial ring operation events occur during daily operation of the VVER unit, but no accurate nuclear thermal coupling calculation analysis is performed on the partial ring operation working condition of the VVER unit at present in China, only the control mode under the working condition is researched, and the main achievement is focused on the core operation level. Disclosure of Invention The invention provides a core physical optimization calculation method under a nuclear power unit partial-ring operation condition, which is used for solving the problem that a nuclear thermal coupling calculation method aiming at the partial-ring operation condition is lacked in the prior art. The technical scheme of the invention is as follows: the invention provides a core physical optimization calculation method under the operating condition of a nuclear power unit deflection ring, which comprises the following steps: Step one, constructing a geometrical model and a reactor core grid model of a reactor in fluid dynamics software; step two, constructing a porous medium model in fluid dynamics software; Setting boundary conditions in fluid dynamics software, carrying out thermodynamic and hydraulic calculation, and carrying out normalization processing on the flow of each fuel assembly of the reactor core through average flow on the flow of the reactor core inlet to obtain reactor core inlet flow distribution; Step four, constructing a thermodynamic and physical coupling interface, taking a thermodynamic calculation result of fluid dynamics software as input, and carrying out physical calculation in the physical calculation software to obtain a reactivity coefficient of a reactor core and three-dimensional power distribution, wherein the three-dimensional power distribution comprises a power peak factor Kq on a fuel assembly and a fuel assembly volume power peak factor Kv; Taking three-dimensional power distribution of physical calculation software as fluid dynamics software input, carrying out thermodynamic and hydraulic calculation again to obtain updated calculated reactor core inlet flow distribution, taking the result as input, and carrying out physical calculation on the physical calculation software again to obtain optimally calculated fuel assembly upper power peak factor Kq and fuel assembly volume power peak factor Kv; and step six, comparing the power peak factor Kq on the fuel assembly and the power peak factor Kv on the fuel assembly in the step four and the step five, and repeating the steps three to five until the deviation of the power peak factor Kq on the fuel assembly and the power peak factor Kv on the fuel assembly meets the precision requirement. In some embodiments, the geometric model of the reactor in the first step comprises a core fuel assembly, a core coaming, a core basket, a fuel assembly supporting seat, a flow isolation ring, a protection pipe assembly, an inner wall surface of a reactor pressure vessel and a coolant flow inlet and outlet area, and after the geometric model is established, the geometric model is subjected to grid division to construct a core grid model. In some embodiments, when the geometric model is built in the first step, the pressure vessel model is adjusted or simplified on the premise of not influencing the integral flow characteristic in the pressure vessel, the ellipsoidal bottom head of the reactor core basket model is equivalent to 1 opening on the premise of ensuring that the coolant uniformly enters the lower cavity, and the middle part of the fuel assembly supporting seat model is equivalent to 6 narrow grooves uniformly distributed along the circumferential direction of the cylinder on the premise of ensuring the flow area. In so