CN114757122A - Method for establishing fine thermal hydraulic calculation model for sodium-cooled fast reactor core disintegration accident - Google Patents
Method for establishing fine thermal hydraulic calculation model for sodium-cooled fast reactor core disintegration accident Download PDFInfo
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Abstract
The invention discloses a method for establishing a fine thermal hydraulic calculation model of a sodium-cooled fast reactor core disintegration accident, which comprises the following steps: 1. carrying out coupling analysis aiming at the behavior characteristics of the reactor core in the initial stage and the transition stage of the reactor core disintegration accident; 2. establishing a three-dimensional compressible two-phase moving particle semi-implicit (MPS) numerical simulation method on a sodium pool; 3. establishing a finite element mechanical model of a material stress-strain constitutive equation; 4. iterative updating of the cross-scale two-phase fluid thermal hydraulic characteristics of the sodium pool scale-narrow gap channel is carried out, and mutual transmission of physical parameters and advanced difference value coupling calculation are realized; 5. establishing a continuous surface tension model, a component transportation model and a heat conduction model; 6. three-dimensional two-phase numerical simulation in the narrow-gap discharge process is realized; 7. establishing a set of two-phase discharge thermal hydraulic mechanism model suitable for a wide parameter range; 8. the cross-scale coupling of the impact-discharge fluid domains under the same time layer is realized, and the coupling change trends of key thermal hydraulic parameters of two phenomena are obtained.
Description
Technical Field
The invention belongs to the technical field of sodium-cooled fast reactor accident safety analysis, and particularly relates to a method for establishing a fine thermal hydraulic calculation model of a sodium-cooled fast reactor core disintegration accident.
Background
As one of the more rapidly and deeply researched reactor types in the fourth generation advanced reactor, the sodium-cooled fast reactor has the advantages of nuclear fuel proliferation, long-life radioactive waste transmutation and inherent safety. The design of the pool type sodium-cooled fast reactor reduces the leakage of the coolant, and the function and the structural integrity of the inclusion boundary are the core problems of the safety characteristic of the pool type sodium-cooled fast reactor. In a reactor primary coolant container, a safety pool surrounded by the outer side of a sodium pool can contain sodium leaked under an accident; the top of the sodium pool contains liquid metal sodium and fission products by means of a shielding cover, so that the evaluation of a bleeding mechanism at the sealing part of the penetrating piece under accidental load is very important.
According to the principle of deep defense, the containment capability of the reactor boundary to radioactive substances under the limit accident of reactor core disintegration is the key point of attention. The analysis method of the reactor core disintegration accident is firstly proposed by Bethe-Tait. Zhandonghui et al improve the B-T model, introduce the distribution function of reactor core reactivity coefficient, and obtain the conservative analysis result. Thilak et al studied the entrained motion characteristics of core air cavity-coolant particles during the expansion of the air cavity. ASTARTE and SEURBNUK respectively calculate the compressible fluid domain based on finite difference method of Lagrange method and Eulerian method, and solve by coupling solid domain finite element method. Louvet al established a scaling experiment MARA10 of French phoenix reactor vessel 1:30 based on MARA 1-9 model experiments, considered core support and upper core structure MARA10, and experimentally verified a finite difference-finite element strongly coupled molten air cavity-sodium-argon three-phase analysis program. The core disintegration and expansion stage research focuses more on the heat exchange phenomenon of a core molten air cavity and the integral effect of a coolant and a reactor vessel, and lacks the high-energy sodium-argon two-phase fluid transient characteristic research and the analysis of a top cover impact failure mechanism coupling a three-dimensional velocity field, a pressure field and a temperature field of the fluid.
Both the finite element method and the finite volume method are numerical methods based on grids, and grid distortion can occur when large-scale liquid level deformation is simulated; the particle method is a Lagrange method for incompressible fluid, and can accurately simulate and track complex liquid level changes. The existing research on the coolant impact top cover focuses on the near-normal pressure effect under the earthquake load, the adopted particle method is a two-dimensional incompressible model, the three-dimensional flow field characteristics of gas phase compression and fluid temperature pressure stress load under high pressure in the expansion stage cannot be analyzed, and the difficulty is brought to the determination of the initial state of two-phase leakage of the coolant after the sealing failure. Chellapandi et al simulated the disassembly expansion process of the indian fast reactor PFBR core using steel vessels with a ratio of 1:13 to 1:30, and obtained the two-dimensional stress-strain characteristics of the vessel by developing the finite element calculation program FUSTIN. The research on the reactor core disintegration and expansion stage focuses more on the heat exchange phenomenon of the reactor core molten air cavity and the integral effect of the coolant and the reactor vessel, and the research on the transient characteristics of high-energy sodium-argon two-phase fluid and the analysis on a top cover impact failure mechanism are lacked.
The research on the impact failure of the reactor vessel and the discharge characteristic of the coolant in the later stage of the reactor core disintegration accident is an important subject of the safety design of the sodium-cooled fast reactor, and the current research mainly has the following problems: first, the existing mature core crash disassembly analysis programs lack systematic investigation of the core expansion phase. Secondly, the impact phenomenon of the coolant on the top cover of the reactor at the stage is complex, the distribution characteristics of the two-phase flow field and the fluid-solid three-dimensional coupling mechanism under multiple physical fields are lack of deep research, and the conventional research method needs to be improved on compressible two-phase interface tracking and fluid transient stress load; thirdly, the research on the sodium-argon two-phase discharge mechanism in the complex narrow-gap channel is deficient, and a model of the narrow-gap channel coolant discharge characteristic under large pressure difference needs to be established.
Disclosure of Invention
In the later stage of reactor core disassembly of the pool type fast reactor, the coolant impacts the reactor vessel to cause failure discharge, the integrity of the second containment boundary is damaged, and the safety of the containment structure is threatened while the cooling capacity of the reactor core is deteriorated. In order to research the formation mechanism of the impact failure and the coolant discharge process, grasp the two-phase flow heat exchange characteristic of the process and reveal the interaction mechanism of the coolant and the primary side boundary in the process. The invention aims to provide a method for establishing a fine thermal hydraulic calculation model of a sodium-cooled fast reactor core disintegration accident, which is used for evaluating the influence of high-energy coolant impact on the structural integrity of a reactor vessel and exploring the discharge characteristics of two phases of liquid metal sodium and inert gas. The method provides theoretical support and basis for safety characteristic research and reactor containment boundary optimization design under the extreme accident of the pool type sodium-cooled fast reactor in China.
In order to achieve the purpose, the invention adopts the following technical scheme:
a method for establishing a fine thermal hydraulic calculation model of a reactor core disassembly accident of a sodium-cooled fast reactor is characterized in that a phenomenon that a coolant impacts a top cover of a sodium-cooled fast reactor to cause seal failure and discharge is used as a background, a moving particle semi-implicit method is improved, three-dimensional solid domain modeling is carried out on an integral large-scale structure of the top cover of the container and a local fine structure of a sealing member, a set of two-phase discharge thermal hydraulic mechanism model suitable for a wide parameter range is established, so that a key thermal hydraulic parameter influence rule in the impact and discharge processes is obtained, and a theoretical basis can be established for evaluating the safety performance of the sodium-cooled fast reactor container and optimizing and designing the reactor structure; the method comprises the following steps:
Step 1: based on a two-dimensional single-phase moving particle semi-implicit MPS method, a pool type sodium-cooled fast reactor geometric model is established, and the model comprises a coolant flowing area in a pool type sodium-cooled fast reactor core, the inner wall surface and the outer wall surface of a pressure vessel, a top shielding cover and a coolant in the reactor core.
And 2, step: modifying an MPS surface tension model on the basis of the step 1, determining transient characteristics of a gas phase interface according to the surface tension, considering gas phase compressibility, adding a gas phase expansion model and a gravity term, then resetting particle arrangement, constructing a local node grid-free convection format in an Euler coordinate system, generating a one-dimensional local node in the flow direction, acquiring next time-layer physical parameters by an interpolation method, and simultaneously limiting parameter extremum of a high-order convection format to stabilize a calculation result;
and 3, step 3: carrying out simplified simulation on the melting and vaporization behaviors of the reactor core in the initial and transition stages, and determining the initial state conditions of the expansion of the air cavity of the reactor core, such as the power, the reactivity, the temperature and the pressure of the reactor core and the like by considering the melting redistribution and the Doppler effect of the reactor core; then establishing a three-dimensional fluid domain model for the molten gas cavity of the reactor core, liquid sodium around and above the molten gas cavity of the reactor core and an inert gas cavity covering a sodium pool, wherein the three-dimensional fluid domain model comprises reactor vessel side walls, top cover wall surfaces, a central measuring column, a pump, a heat exchanger and other penetrating components; and finally, carrying out numerical calculation to obtain three-dimensional temperature and pressure field, velocity field and phase state distribution characteristics of the liquid metal sodium-argon two-phase fluid under expansion and high pressure during impact of the coolant, and finely capturing phase interface transient behavior in the flow field.
And 4, step 4: according to transient characteristics of a three-dimensional velocity field, a pressure field and a temperature field of two-phase fluid in the impact process, mechanical performance analysis of a reactor structure is carried out, a finite element mechanical model of a material stress-strain constitutive equation is established, three-dimensional solid domain modeling is respectively carried out on a whole large-scale structure of a top cover of the reactor and a local fine structure of a sealing member, and fluid-solid coupling research under stress load of multiple physical fields in the impact process is realized.
And 5: and (3) considering a local fluid region of a coolant discharge gap formed by sealing failure, performing iterative updating of the cross-scale two-phase fluid thermal hydraulic characteristics of the sodium pool scale-narrow gap channel, establishing a coupling surface displacement model and realizing mutual transmission of physical parameters and advanced difference value coupling calculation.
And 6: under the condition of considering multiple two-phase flow states, complex gap morphology and other factors, a continuous surface tension model, a component transportation model and a heat conduction model are established, and three-dimensional two-phase numerical simulation is carried out on a narrow gap release process through the coupled solution of a volume function model, a primary crushing model and a discrete phase model.
And 7: on the basis of numerical analysis of the two-phase discharge characteristic, a set of two-phase discharge thermal hydraulic mechanism model suitable for a wide parameter range is established, and the two-phase discharge thermal hydraulic mechanism model is verified by adopting a numerical analysis result. And transferring the calculation result of the two-phase discharge thermodynamic and hydraulic mechanism model to an impact three-dimensional flow field calculation domain to realize cross-scale coupling of impact-discharge fluid domains under the same time layer, and finally obtaining a narrow-gap thermodynamic and hydraulic mechanism model and discharge rate prediction.
The method for establishing the three-dimensional fine thermal hydraulic calculation model under the sodium-cooled fast reactor core disintegration accident is suitable for the research of the sodium-cooled fast reactor characteristics under the extreme accident, and compared with the existing research method, the method has the following innovative characteristics:
1. according to the invention, in a two-dimensional single-phase moving particle semi-implicit Method (MPS), a moving particle semi-implicit method suitable for a compressible two-phase flow field is developed by correcting a surface tension model and considering conditions such as gas phase compressibility, and compared with the traditional MPS method, the distribution characteristics of the flow field in the later stage of a reactor core disintegration accident under a three-dimensional field in a coolant impact process can be effectively disclosed;
2. according to transient characteristics of a two-phase fluid three-dimensional velocity field, a pressure field and a temperature field in an impact process, the invention establishes a cross-scale three-dimensional fluid-solid coupling analysis method under a multi-physical field so as to deeply explain the interaction between the thermal hydraulic characteristics and the mechanical characteristics of materials of a reactor in the process of impacting a top cover by a high-speed coolant and disclose a primary side coolant boundary failure mechanism.
3. The method obtains the thermal hydraulic characteristics of the discharge process by coupling solution of the volume function model, the primary crushing model and the discrete phase model under the condition of considering factors such as multiple two-phase flow state, complex gap morphology and the like, develops a cross-scale parameter coupling transfer method of the flow field and explains the change rule of key parameters of the impact and discharge processes.
Drawings
FIG. 1 is a flow chart of the method of the present invention.
FIG. 2 is a process schematic diagram of a high-energy coolant impact three-dimensional flow field thermal hydraulic characteristic research part under a pool type sodium-cooled fast reactor large span;
FIG. 3 is a schematic diagram of a failure mechanism analysis process of a reactor vessel based on bidirectional fluid-solid coupling;
FIG. 4 is a schematic diagram of a multi-physical-field cross-scale thermo-mechanical bidirectional fluid-solid coupling method;
fig. 5 is a schematic diagram of a narrow-gap channel two-phase discharge multi-flow heat transfer characteristic research process.
Detailed Description
The invention is described in further detail below with reference to the drawings of the specification:
as shown in FIG. 1, the invention provides a method for establishing a fine thermal hydraulic calculation model of a sodium-cooled fast reactor core disintegration accident, which is divided into three parts. FIG. 2 shows the calculation part of the thermal hydraulic characteristics of the high-energy coolant impact three-dimensional flow field under the large span of the pool type sodium-cooled fast reactor. The method comprises the following steps:
step 1: establishing a pool type sodium-cooled fast reactor geometric model based on a two-dimensional single-phase moving particle semi-implicit MPS (MPS), wherein the model comprises a coolant flowing area in a pool type sodium-cooled fast reactor core, the inner wall surface and the outer wall surface of a pressure vessel, a top shielding cover and a coolant in the reactor core; the basic MPS implementation in fig. 2 is: according to the model, the process that the gas cavity is expanded sharply from the initial stage to the expansion stage is simulated, and the upper liquid sodium metal impacts the top cover at a high speed to cause the sealing failure of the reactor container is accelerated. Finally, obtaining the thermal-hydraulic behavior characteristics of crushing, fusion, falling back and the like of the coolant in the reactor core through an MPS program;
And 2, step: on the basis of the step 1, correcting an MPS surface tension model, determining transient characteristics of a gas phase interface, considering gas phase compressibility, adding a gas phase expansion model and a gravity term, then resetting particle arrangement in a speed divergence mode, constructing a local node grid-free convection format in an Euler coordinate system, generating one-dimensional local nodes in the flow direction, obtaining next time-layer physical parameters through an interpolation method, and simultaneously limiting parameter extreme values of a high-order convection format to stabilize a calculation result. In order to derive a compressible mobile fluid semi-implicit method, comparing the obtained calculation result with a Burgers equation I and a two-dimensional flow standard calculation example analytic value to verify a model, and further verifying the accuracy of the calculation method by adopting bubble growth and collapse experimental data;
and 3, step 3: on the basis of the step 2, carrying out simplified simulation on the melting and vaporization behaviors of the reactor core in the initial and transition stages, and determining the power, reactivity and temperature and pressure distribution of the reactor core by considering the melting redistribution and Doppler effect of the reactor core so as to obtain the initial mass energy source item of the molten material air cavity of the reactor core; then establishing a three-dimensional fluid domain for the molten gas cavity of the reactor core, liquid sodium around and above the molten gas cavity of the reactor core and an inert gas cavity covering a sodium pool, wherein the three-dimensional fluid domain comprises accurate modeling of penetrating components such as the side wall and the top cover wall surface of the reactor container, a central measuring column, a pump, a heat exchanger and the like; finally, numerical calculation is carried out on the basis of the first two steps to obtain three-dimensional temperature and pressure fields, velocity fields and phase state distribution characteristics of the liquid metal sodium-argon two-phase fluid in the process of impacting the coolant under expansion and high pressure, phase interface transient behaviors in the flow field are captured finely, and a gas phase compression entrainment mechanism is analyzed;
And 4, step 4: and (3) obtaining the distribution characteristics of the three-dimensional flow field of the coolant under the expansion high pressure in the sodium-cooled fast reactor container through the steps 1, 2 and 3. Based on the transient characteristics of the three-dimensional temperature and pressure field and the velocity field of the two-phase fluid in the obtained calculation result, a structural mechanical model of a finite element method is established, three-dimensional solid domain modeling is respectively carried out on the whole large-scale structure of the top cover of the reactor container and the local fine structure of the sealing member, and fluid-solid coupling research under the stress load of multiple physical fields in the impact process is realized;
and 5: on the basis of the step 4, the iterative updating of the thermal hydraulic characteristics of the trans-scale two-phase fluid of the sodium pool scale-narrow gap channel is carried out by considering the local fluid area of the coolant discharge gap formed by the seal failure (figure 3), and the specific method is shown in figure 4: and (4) selecting FLUENT, MPCCI and Abaqus tools to analyze and solve the multi-physical-field interaction process. The fluid domain adopts a finite volume method, and the solid domain adopts a finite element method. And separately solving the mass, momentum and energy conservation equations of the fluid domain and the elastoplasticity calculation model of the solid domain under the same time-space layer. Carrying out secondary development on the two-phase flow model and the dynamic grid model in the fluid domain; and in the solid domain, performing secondary development on an elastic-plastic model and a strong nonlinear contact model. Obtaining an iteration result, realizing a simulation verification technology combining multi-source data and multiple models, establishing a coupling surface displacement model and realizing mutual transmission of physical parameters and advanced difference value coupling calculation based on the developed impact two-phase numerical simulation and two-phase discharge mechanism characteristic research;
And 6: the analysis results of the step 3 and the step 5 provide initial fluid phase state conditions and flow channel structure conditions for the discharge process research at different positions. Under the condition of considering multiple two-phase flow state, complex gap morphology and other factors, establishing a continuous surface tension model, a component transportation model and a heat conduction model, and performing three-dimensional two-phase numerical simulation on a narrow-gap discharge process through the coupled solution of a volume function model, a primary crushing model and a discrete phase model, wherein the three-dimensional two-phase numerical simulation specifically comprises the following steps: firstly, two-phase three-dimensional numerical simulation research is carried out on the leakage flow phenomenon of the sealing clearance of typical top cover penetration pieces such as a pump, a heat exchanger, a rotary shielding plug and the like. Determining boundary conditions according to a three-dimensional two-phase flow field in an impact process; and carrying out gradient space discretization on a flow region in the gap, selecting a k-omega two-equation model, adding a gravity and surface tension correction model, and carrying out numerical simulation on the bleeding phenomenon by adopting a second-order windward format discrete momentum, energy and turbulence relation. The interaction of the argon and the liquid sodium layer influences the distribution of two flowing phases in the gap, and a gas component transport equation is introduced; aiming at a multiple two-phase flow pattern from a dispersion flow to a bubble flow, establishing a coupling solving method of a fluid volume function (VOF) model, a primary fragmentation model (PBM) and a Discrete Phase Model (DPM);
And 7: as shown in fig. 5, on the basis of numerical analysis of the narrow-gap bleeding process, a set of two-phase bleeding thermodynamic and hydraulic mechanism model suitable for a wide parameter range is established: and analyzing a discharge mechanism phenomenon based on a numerical simulation result, and obtaining a mechanism relation of key parameters of the model according to the special flow channel morphology of typical gaps, small opening width, large section length-width ratio and rough distortion, thereby developing flow resistance characteristic models of single-phase and two-phase flow inlets, wall surfaces, corner friction and the like. And transferring the model calculation result to an impact three-dimensional flow field calculation domain to realize the cross-scale coupling of the impact-discharge fluid domain under the same time layer, and finally obtaining a narrow-gap thermodynamic and hydraulic mechanism model and discharge rate prediction.
Claims (1)
1. A method for establishing a fine thermal hydraulic calculation model of a sodium-cooled fast reactor core disintegration accident is characterized by comprising the following steps: on the background of the phenomenon that the coolant impacts the top cover of the sodium-cooled fast reactor container to cause seal failure and therefore discharge, a moving particle semi-implicit method is improved, three-dimensional solid domain modeling is carried out on the whole large-scale structure of the top cover of the container and the local fine structure of a sealing member, a set of two-phase discharge thermal hydraulic mechanism model suitable for a wide parameter range is established, so that the influence rule of key thermal hydraulic parameters in the impact and discharge processes is obtained, and a theoretical basis can be laid for evaluating the safety performance of the sodium-cooled fast reactor container and optimizing and designing the structure of a reactor;
The method comprises the following steps:
step 1: establishing a geometric model of the pool type sodium-cooled fast reactor based on a two-dimensional single-phase moving particle semi-implicit method MPS, wherein the model comprises a coolant flowing area in the pool type sodium-cooled fast reactor core, the inner wall surface and the outer wall surface of a pressure vessel, a top shielding cover and coolant in the reactor core;
and 2, step: modifying an MPS surface tension model on the basis of the step 1, determining transient characteristics of a gas phase interface according to the surface tension, then resetting particle arrangement, generating one-dimensional local nodes in the flow direction by constructing a local node grid-free convection format in an Euler coordinate system, acquiring next time-layer physical parameters by an interpolation method, and simultaneously limiting parameter extremum of a high-order convection format to stabilize a calculation result;
and step 3: carrying out simplified simulation on the melting and vaporization behaviors of the reactor core in the initial and transition stages, and determining the expansion initial state condition of the air cavity of the reactor core by considering the melting redistribution and Doppler effect of the reactor core; then establishing a three-dimensional fluid domain model for the reactor core molten gas cavity, liquid sodium around and above the reactor core molten gas cavity and an inert gas cavity covering the sodium pool, wherein the model comprises a side wall and a top cover wall surface of the sodium-cooled fast reactor container, a central measuring column, a pump and a heat exchanger; finally, carrying out numerical calculation to obtain three-dimensional temperature and pressure fields, velocity fields and phase state distribution characteristics of the liquid metal sodium-argon two-phase fluid in the process of impacting the coolant under the expansion high pressure, and finely capturing the phase interface transient behavior in the flow field;
And 4, step 4: according to transient characteristics of a three-dimensional temperature and pressure field and a velocity field of the two-phase fluid in the impact process, mechanical performance analysis of a reactor container structure is carried out, three-dimensional solid domain modeling is respectively carried out on a whole large-scale structure of a top cover of the reactor container and a local fine structure of a sealing member, and fluid-solid coupling research under stress load of multiple physical fields in the impact process is realized;
and 5: considering a local fluid region of a coolant discharge gap formed by sealing failure, performing iterative update of the cross-scale two-phase fluid thermal hydraulic characteristics of the sodium pool scale-narrow gap channel, establishing a coupling surface displacement model and realizing mutual transmission of physical parameters and advanced difference value coupling calculation;
step 6: under the condition of considering multiple two-phase flow states, complex gap morphology and other factors, establishing a continuous surface tension model, a component transportation model and a heat conduction model, and performing three-dimensional two-phase numerical simulation on a narrow-gap discharge process through the coupled solution of a volume function model, a primary crushing model and a discrete phase model;
and 7: on the basis of numerical analysis of a narrow-gap discharge process, establishing a set of two-phase discharge thermal hydraulic mechanism model suitable for a wide parameter range, verifying the two-phase discharge thermal hydraulic mechanism model by adopting a numerical analysis result and carrying out discharge rate prediction research under multiple factors; and transferring the calculation result of the two-phase discharge thermal hydraulic mechanism model to an impact three-dimensional flow field calculation domain to realize cross-scale coupling of impact-discharge fluid domains under the same time layer, and finally obtaining a narrow-gap thermal hydraulic mechanism model and discharge rate prediction.
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Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN115270660A (en) * | 2022-08-04 | 2022-11-01 | 上海交通大学 | Multi-scale multi-physical field coupling analysis method for transient behavior of space thermionic reactor |
| CN115688488A (en) * | 2022-12-30 | 2023-02-03 | 中国核动力研究设计院 | Criterion determination method, criterion determination device, criterion determination equipment, computer storage medium and program product |
Citations (7)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH0682592A (en) * | 1992-07-06 | 1994-03-22 | General Electric Co <Ge> | Automatic core monitoring apparatus |
| US20150243376A1 (en) * | 2014-02-26 | 2015-08-27 | Taylor Ramon WILSON | Molten salt fission reactor |
| CN110532586A (en) * | 2019-07-08 | 2019-12-03 | 华北电力大学 | A kind of the subregion decoupling modeling and overall coupling calculation of sodium-cooled fast reactor container |
| CN110543705A (en) * | 2019-08-19 | 2019-12-06 | 西安交通大学 | Boiling simulation solving acceleration method in typical channel of nuclear reactor |
| CN110867220A (en) * | 2019-11-07 | 2020-03-06 | 西安交通大学 | Method for researching in-core eutectic reaction and high-temperature melting behavior by particle grid mixing method |
| CN111832214A (en) * | 2020-06-29 | 2020-10-27 | 西安交通大学 | Method for simulating melting process of nuclear reactor severe accident scrap bed based on particle method |
| CN113221480A (en) * | 2021-05-14 | 2021-08-06 | 西安交通大学 | Failure analysis method for reactor vessel in sodium-cooled fast reactor core disassembly accident |
-
2022
- 2022-04-19 CN CN202210407443.7A patent/CN114757122B/en active Active
Patent Citations (7)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH0682592A (en) * | 1992-07-06 | 1994-03-22 | General Electric Co <Ge> | Automatic core monitoring apparatus |
| US20150243376A1 (en) * | 2014-02-26 | 2015-08-27 | Taylor Ramon WILSON | Molten salt fission reactor |
| CN110532586A (en) * | 2019-07-08 | 2019-12-03 | 华北电力大学 | A kind of the subregion decoupling modeling and overall coupling calculation of sodium-cooled fast reactor container |
| CN110543705A (en) * | 2019-08-19 | 2019-12-06 | 西安交通大学 | Boiling simulation solving acceleration method in typical channel of nuclear reactor |
| CN110867220A (en) * | 2019-11-07 | 2020-03-06 | 西安交通大学 | Method for researching in-core eutectic reaction and high-temperature melting behavior by particle grid mixing method |
| CN111832214A (en) * | 2020-06-29 | 2020-10-27 | 西安交通大学 | Method for simulating melting process of nuclear reactor severe accident scrap bed based on particle method |
| CN113221480A (en) * | 2021-05-14 | 2021-08-06 | 西安交通大学 | Failure analysis method for reactor vessel in sodium-cooled fast reactor core disassembly accident |
Non-Patent Citations (2)
| Title |
|---|
| 刘余;张虹;贾宝山;: "核反应堆热工水力多尺度耦合模拟初步研究", 核动力工程, no. 1, 25 May 2010 (2010-05-25) * |
| 贾斌;马帅;史强;高新力;靖剑平;石兴伟;: "非能动压水堆热工水力多尺度耦合计算分析研究", 核科学与工程, no. 05, 15 October 2018 (2018-10-15) * |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN115270660A (en) * | 2022-08-04 | 2022-11-01 | 上海交通大学 | Multi-scale multi-physical field coupling analysis method for transient behavior of space thermionic reactor |
| CN115688488A (en) * | 2022-12-30 | 2023-02-03 | 中国核动力研究设计院 | Criterion determination method, criterion determination device, criterion determination equipment, computer storage medium and program product |
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