Luận án tiến sĩ cơ học nghiên cứu hiện tượng chuyển pha trong vùng hoạt lò phản ứng

TABLE OF CONTENTS Abbreviations and Nomenclature .8 List of Tables 12 List of Figures .14 Overview 17 Chapter Introduction to research work .19 1.1 Status of nuclear power in the World and Vietnam 19 1.2 Brief overview of nuclear safety 20 1.3 Core thermal hydraulics safety analysis in transient condition 21 1.3.1 Role of void fraction in simulation of two phase flow 24 1.3.2 Experiment overview for bundle of sub channel analysis 25 1.3.3 Void fraction prediction study .26 1.4 VVER technology understanding related to this study 27 1.5 Thesis objectives 29 1.5.1 Studied object 30 1.5.2 Scope of study .30 1.6 Thesis outline .31 Chapter Overview of phase change models in code theories with different scales 33 2.1 Multi code and multi scales approach to PWR thermal hydraulic simulation 33 2.1.1 Neutron codes and thermal hydraulics codes 33 2.1.2 Different scale of thermal hydraulic codes 34 2.1.3 Different thermal hydraulic modeling approaches 36 2.2 Phase change models in system code RELAP5 38 2.3 Phase change models in sub channel code CTF 40 2.3.1 Evaporation and condensation induced by thermal phase change 40 2.3.2 Evaporation and condensation induced by turbulent mixing and void drift 42 2.4 Phase change models in meso scale code CFX 42 2.4.1 Evaporation at the wall 42 2.4.2 Condensation model in bulk of liquid 43 2.5 Conclusions 44 Chapter Phase change models verification and assessment by numerical simulation .45 3.1 Brief information of VVER-1000/V392 45 3.2 Verification of RELAP5 simulation models for VVER-1000/V392 reactor with SAR 47 3.2.1 Nodalization scheme 48 3.2.2 Verification of modeling through steady-state study 48 3.2.3 Verification through accident case study 49 3.3 CTF models verification and assessment with BM ENTEK tests 51 3.3.1 ENTEK BM facility 51 3.3.2 Modeling by CTF 53 3.3.3 Results and discussions .53 3.4 Verification CFX models with PSBT sub channel tests 59 3.4.1 PSBT test section for single sub channel 60 3.4.2 Mesh generation study 61 3.4.3 Solver convergence study 63 3.4.4 Mesh refinement study 64 3.4.5 Sensitivity study on physical models 68 3.4.6 Assessment of CFX and CTF modeling results in comparison with PSBT single channel 79 3.4.7 Discussion on CTF and CFX void fraction predictions 82 3.4.8 Improvement of CFX void fraction prediction in saturated region .84 3.5 Conclusions 86 Chapter Void fraction prediction in hot channel of VVER-1000/V392 88 4.1 Calculation Diagram 88 4.2 Power distribution calculation by MCNP5 code 90 4.3 LOCAs simulation by RELAP5 code 93 4.4 Void fraction prediction in hot channel during transient by CTF code 96 4.4.1 VVER-1000/V392 void fraction prediction by CTF 96 4.4.2 Discussion on RELAP5 and CTF void fraction predictions .98 4.5 Void fraction prediction in single channel by CFX code 100 4.5.1 Mesh refinement study 101 4.5.2 Void fraction prediction calculated by CFX along sub channel 102 4.6 Void fraction prediction in bundle of channel calculated by CFX code 104 4.7 Conclusions 107 Conclusions and proposals 108 Achievements and new findings given by the thesis 108 Proposal of future work 110 References 112 List of Author’ papers and report 116 Abbreviations and Nomenclature Abbreviations VVER VVER-1200/V491 VVER-1000/V392 VINATOM TSO DID PWR SAR NRA RIAs LOFAs LOCAs DNB DNBR Castellana EPRI BM ENTEK RBMK-1000 PSBT CTF RELAP5 COBRA-TF RELAP-3D MARS-3D Belene Ansys CFX CFX PARCS ITT 0D, 1D, 2D CHF TH RANS A Type of Pressurized Water Reactor developed by Russia A type of Russia reactor with capability of 1200 MWe A type of Russia reactor with capability of 1000 MWe Vietnam Atomic Energy Institute Technical Support Organization Defend in depth policy in nuclear power plant design Pressurized Water Reactor Safety Analysis Report of nuclear power plant Nuclear Regulatory Authority Reactivity insertion accident Loss of coolant flow Loss of coolant accident Departure of nucleate boiling Departure of nucleate boiling ratio The x square rod bundle test for fuel rod in Columbia University (USA) Electric Power Research Institute The BM Facility at the Research and Development Institute of Power Engineering (RDIPE; a.k.a., ENTEK and NIKIET) models the forced circulation circuit of RBMK type reactors A type of Russia reactor of 1000 MWe with transliteration of Russian characters for graphite-moderated boiling-water-cooled channel-type reactor OECD/NRC Benchmark based on Nuclear Power Engineering Corporation (NUPEC, Japan) PWR sub channel and bundle tests A version of COBRA-TF improved by Pennsylvania State University (USA) System code developed by Information Systems Laboratories, Inc Rockville, Maryland Idaho Falls, Idaho Coolant-Boiling in Rod Arrays—Two Fluids (COBRA-TF) is a Thermal Hydraulic (T/H) simulation code designed for Light Water Reactor (LWR) vessel analysis developed by Pacific Northwest Laboratory Newest version of RELAP5 with coupling with COBRA-TF Newest version of MARS with coupling with COBRA-TF A site for nuclear power plant project in Bulgaria A Computational Fluid Dynamics developed by Ansys Same as Ansys CFX A code for neutron kinetic calculation interface tracking technique Dimension of spatial averaging Critical Heat Flux Thermal hydraulics Reynolds-averaged Navier–Stokes Simulation LES MSLB PTS CFD DI FI SI U-RANS T-RANS meso scale ECCS system LBLOCAs SBO SG SG PHRS HA-2 HA-1 PCT DBA MCPL LOOP DG SAR SG OECD/NRC BFBT αcrit Large Eddy Simulation Main steam line break Pressurize Thermal shock Computational Fluid Dynamics Deterministic Interface Filtered Interface Statistical Interface Unsteady flow Transient flow The spatial scale with size around 1mm and less simulated with RANS Emergency Core Cooling System Large break for loss of coolant accident Station black out Steam Generator Passive Heat Removal through Steam Generator Secondary stage of Hydro accumulators First stage of Hydro accumulators Peaking temperature of cladding Design Base Accident Main Coolant Pipe line Loss of offsite power Diesel Generator SG Active Heat Removal System UPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark Void fraction corresponding with critical heat flux correlation Nomenclature As Ax Cpl Cpv ̅ hc hl* hg ̇ ̇ Ρl Qwf Qwif, Qboil Tg TS Tcrit Tl, Tf rb ̅ Γ’’’ Sub-cooled vapor interfacial area per unit volume (m-1) Super-heated liquid interfacial area per unit volume (m-1) Super-heated vapor interfacial area per unit volume (m-1) Conductor surface area in mesh cell (m2) Mesh-cell area, X normal (m2) Liquid specific heat, constant pressure (J/kg.K) Vapor specific heat, constant pressure (J/kg.K) Mixing mass flux (kg/m2.s) Vapor saturation enthalpy (J/kg) Sub-cooled liquid interface heat transfer coefficient (W/m2.K) Sub-cooled vapor interface heat transfer coefficient (W/m2.K) Super-heated liquid interface heat transfer coefficient (W/m2.K) Super-heated vapor interface heat transfer coefficient (W/m2.K) Chen correlation heat transfer coefficient (W/m2.K) Liquid enthalpy (J/kg) Liquid saturation enthalpy (J/kg) Vapor enthalpy (J/kg) Vapor interface heat transfer coefficient (W/m3 K) Liquid interface heat transfer coefficient (W/m3 K) Mass exchange due to drift model (kg/s) Mass exchange of phase k (kg/m2.s) Density of liquid (kg/m3) Wall heat transfer to liquid (W) Wall heat transfer to liquid for convection (W) Wall heat transfer to liquid for vaporization (W) Vapor temperature (K) Saturated temperature (K) Critical heat flux temperature (K) Liquid temperature (K) Bubble diameter (m) Void fraction of phase k induced by sub channel i Equilibrium quality void fraction Two phase turbulent mixing coefficient Density of phase k in sub channel i (kg/m3) Liquid density (kg/m3) Vapor density (kg/m3) Mixing density (kg/m3) Volumetric mass flow rate (kg/m3.s) Vapor generation from near wall (kg/m3.s) Total Vapor Generation (kg/m3.s) Mesh-cell axial height (m) Surface tension (N/m) Fluid viscosity (Pa.s) Pressure (Pa) 10 Γ’’ Tw Tchf ,Tcrit Re Pr Nu n kl , hv hnb hl hg hfc hf hc g FChen f Dh Cp Ax As Fo ̅ ̅ , Evaporation rate (kg/m2.s) Wall surface temperature (K) Critical heat flux temperature (K) Reynolds number Prandtl number Nusselt number Wall nucleation site density (m-2) Liquid thermal conductivity (W/m.K) Vapor enthalpy (J/kg) Nucleate-boiling heat transfer coefficient (W/m2.K) Liquid enthalpy (J/kg) Vapor saturation enthalpy (J/kg) Forced-convective heat transfer coefficient (W/m2.K Liquid saturation enthalpy (J/kg) Chen correlation heat transfer coefficient (W/m2.K) Gravitational acceleration (m/s2) Chen Reynolds number factor Bubble detachment frequency (s-1) Hydraulic diameter (m) Specific heat, constant pressure (J/kg.K) Mesh-cell area, X normal (m2) Conductor surface area in mesh cell (m2) Mesh-cell axial height (m) Inverse Martinelli factor Liquid density (kg/m3) Fourier number Vapor density (kg/m3) Mixing density (kg/m3) Void fraction Volumetric heat transfer from the wall (W/m3) Total wall heat flux (W/m2) Quenching heat flux (W/m2) Evaporative heat flux (W/m2) Convective heat flux (W/m2) Local mean bubble diameter (m) Saturation temperature (K) Liquid temperature (K) Mesh-cell area of phase k (m2) Chen suppression factor Heat transfer per volumetric unit (W/m3) Mixing mass flux (kg/m3.s) Area influence factors 11 List of Tables Table 1.1 Multiple levels of protection from DID approach (source [45]) 20 Table 1.2 Content of Safety Analysis Reports (source [45]) 21 Table 1.3 Castellana 4x4 test characteristics (source [29]) 25 Table 1.4 EPRI 5x5 characteristics for test 74 and test 75(source [29]) 25 Table 1.5 Geometry and power shape for Test Assembly B5, B6, and B7 (Source [1]) .25 Table 2.1 Main characteristics of codes with four different scales (source [11]) 36 Table 2.2 Main characteristics modeling approaches for three main types of single-phase CFD 37 Table 3.1 Main technical characteristics of VVER-1000/V392 (source[36]) 46 Table 3.2 Comparison of steady-state of VVER-1000/V392 .48 Table 3.3 Boundary conditions for event number (source [35]) 49 Table 3.4 Chronological sequence of Event from SAR [35] and this study .50 Table 3.5 Setting for base case and sensitivity cases according to test 01 and test 17 53 Table 3.6 Base case void fraction distribution calculations versus experiment for cases at 3MPa 54 Table 3.7 Base case void fraction distribution calculations versus experiment for cases at 7MPa 54 Table 3.8 Deviation of void fraction distribution calculation results versus experiment .55 Table 3.9 Deviation of void fraction distributions on input uncertainties 58 Table 3.10 Maximum deviation of void fraction distribution on input parameters versus base case 58 Table 3.11 Experimental uncertainties on input parameters 60 Table 3.12 Test Conditions for Steady-State Void Measurement of selected runs 61 Table 3.13 Mesh characteristics .62 Table 3.14 y+ predicted by Mesh .62 Table 3.15 y+ predicted by Mesh .62 Table 3.16 y+ predicted by Mesh .63 Table 3.17 Two phase flow model setting .63 Table 3.18 Average void fraction calculations between three meshes and experiment value .64 Table 3.19 Radial distribution of pressure and temperature for different refinement meshes .64 Table 3.20 Radial distribution of velocity and void fraction for different refinement meshes 65 Table 3.21 Average void fraction calculation at given cross section with or without modeling 69 Table 3.22 Calculation results of average void fraction 73 Table 3.23 Average void fraction calculation with different scale of bubble mean diameter 75 Table 3.24 Average void fraction calculation results with different Nref 77 Table 3.25 Average void fraction calculation results with different bubble departure diameters 78 12 Table 3.26 Average void fraction calculation results with different Nusselt number correlations 79 Table 3.24 CFX and CTF results comparisons versus experiment void fraction 80 Table 3.25 Comparison of CFX and CTF results and experiment void fraction in saturated region 81 Table 3.26 Comparison of CFX and CTF results versus experiment in case of high pressure 81 Table 3.27 Comparison of CFX and CTF results and experiment void fraction 82 Table 3.28 Void fraction and temperature super heating before and after calibration 85 Table 4.1 Main technical characteristics of fuel assembly for VVER-1000/V392 90 Table 4.2 Case studies for void fraction prediction 94 Table 4.3 Boundary condition of LOCA coupled with SBO for analysis 94 Table 4.4 Data related to phase change of interfacial area for case LB01002B at 15s of transient .99 Table 4.5 Cases for void fraction prediction in single channel by CFX 101 Table 4.6 Average void fraction for different meshes 102 Table 4.7 Void fraction prediction by CTF and CFX at downstream of channel at z = 3.48m 102 Table 4.8 Sub cooled selected regions for CFX investigation 105 Table 4.9 Saturated selected regions for CFX investigation 105 13 List of Figures Figure 1.1 Nuclear power generation by country in 2013 (source [46]) 19 Figure 1.2 Multiple physical barriers in DID policy (source [45]) 22 Figure 1.3 Heal flux versus temperature difference for pool boiling heat transfers (source [31]) 23 Figure 1.4 Types of boiling flow crisis (source [25]) 23 Figure 1.5 Critical heat flux in uniformly core (source [25]) 24 Figure 1.6 Development of VVER nuclear reactor technology chart [32] .28 Figure 1.7 Multi-scale analysis of reactor thermal hydraulics (source [11]) 29 Figure 1.7 (a) temperature distributions in a cylindrical fuel pin, (b) flow regime 31 Figure 2.1 Relations between MCNP5, system code RELAP5 and component code CTF 33 Figure 2.3 System code capabilities for reactor thermal hydraulics (source [7]) 34 Figure 2.4 Control volume and axial flow area defined in sub channel code 35 Figure 2.6 The tree of two-phase thermal hydraulic modeling approaches (source [11]) 38 Figure 2.7 Schematic of vertical flow regime map in RELAP5(source [19]) 40 Figure 2.8 CTF normal-wall flow regime maps (source [38]) .41 Figure 3.1 Side view of primary system of VVER-1000/V392 (source [36]) .46 Figure 3.2 Primary system and safety system for VVER-1000/V392 (source [37]) 47 Figure 3.3 VVER-1000/V392 nodalization schemes in this study .48 Figure 3.4 (a) Cladding temperature from calculations, (b) Cladding temperature from SAR 51 Figure 3.5 Test section (Heat Release Zone, φ is diameter in mm) .52 Figure 3.6 BM ENTEK modeling by CTF 53 Figure 3.7 Radial void distribution of the test T04 56 Figure 3.8 Cross mass flow due to turbulent mixing and void drift .56 Figure 3.9 Maximum and minimum voiding curves versus experiment 57 Figure 3.10 Maximum and minimum voiding curves versus experiment 57 Figure 3.11 Uncertainty void fraction distributions for test T01 58 Figure 3.12 Test section for central sub channel void distribution measurement (source [1]) 60 Figure 3.13 Cross section of three proposed meshes .62 Figure 3.14 Base line for radial distribution investigation .64 Figure 3.15 S14326 Radial distribution of void fraction 66 Figure 3.16 S16222 Radial distribution of void fraction 66 Figure 3.17 S12211 Radial distribution of void fraction 67 Figure 3.18 S14326 Axial sub channel distribution of void fraction .67 14 Figure 3.19 S16222 Axial sub channel distribution of void fraction .68 Figure 3.20 S12211 Axial sub channel distribution of void fraction .68 Figure 3.21 S14326 Radial distribution of void fraction of full sub models 70 Figure 3.22 S144411 Radial distribution of void fraction of full sub models 70 Figure 3.23 S12211 Radial distribution of void fraction of full sub models 71 Figure 3.24 S14411 Radial distribution of void fraction of full sub models 71 Figure 3.25 S16222 Radial distribution of void fraction of full sub models 72 Figure 3.26 S14326 Radial distribution of void fraction of full sub models 72 Figure 3.27 S11222 Radial distribution of void fraction with different turbulent 73 Figure 3.28 S16222 Radial distribution of void fraction with different turbulent 74 Figure 3.29 S14326 Radial distribution of void fraction with different turbulent 74 Figure 3.30 S12211 Radial distribution of void fraction with different scale 75 Figure 3.31 S16222 Radial distribution of void fraction with different scale 76 Figure 3.32 S14326 Radial distribution of void fraction with different scale 76 Figure 3.33 (a) Bubble departure size, (b) Heat flux partition with different models 79 Figure 3.34 Temperature distribution along axial and radial channel 84 Figure 3.35 Temperature superheating and void fraction before and after calibration 86 Figure 4.1 Two-phase thermal hydraulic modeling for RELAP5, CTF and CFX 88 Figure 4.2 Geometry of sub channel in VVER-1000/V392 fuel assembly 89 Figure 4.3 VVER-1000/V392 void fraction prediction chart using multi codes and multi scales 90 Figure 4.4 The sixth of core loading pattern and whole core geometry for MCNP5 simulation 91 Figure 4.5 Relative power distribution in the sixth of the whole core 92 Figure 4.6 Distribution of relative power along axial hot channel 93 Figure 4.7 Distribution of relative power in the hot channel 93 Figure 4.8 (a) Whole fuel assembly simulated as hot channel and (b) the active part 95 Figure 4.9 Average void fraction calculated by RELAP5 on exit of active region in hot channel 95 Figure 4.10 Taken twelfth of whole bundle for void fraction prediction 96 Figure 4.11 Cross section of CTF modeling for the selected part of the whole bundle 97 Figure 4.12 Void fraction prediction by CTF and RELAP5 for LBLOCAs 97 Figure 4.13 Void fraction prediction by CTF and RELAP5 for SBLOCAs 98 Figure 4.14 Total vapor generation rate and vapor generation rate near wall 98 Figure 4.15 Total vapor generation rate and vapor generation rate near wall 99 Figure 4.16 Three meshes used to simulate geometry of single channel .101 Figure 4.17 Average void fraction along channel with different meshes 102 15 Figure 4.17 Average void fraction along channel with different meshes for cases (SB01003-09-37) and (SB01003-14-34) The average void fractions calculated at downstream (with z = 3.48m) are nearly the same in all three cases as presented in Table 4.6 Therefore, the mesh M1 is used to investigate the remaining cases in Table 4.5 Table 4.6 Average void fraction for different meshes Case ID Void of mesh M1 at 3.48m Void of mesh M2 at 3.48m Void of mesh M3 at 3.48m SB01003-09-37 SB01003-14-34 SB01003-20-15 0.4064 0.1964 0.2195 0.4093 0.1960 0.2185 0.4074 0.1961 0.2186 4.5.2 Void fraction prediction calculated by CFX along sub channel Table 4.7 show the void fraction prediction in corresponding channel by CTF and CFX The columns named “CTF bundle” and “CTF single” denote for results taken from simulated in all twelfth of bundle or in single sub channel, respectively The difference of void fraction between “CTF bundle” and “CTF single” is caused by turbulent mixing and void drift models in formulas (2.18) and (2.19) For the “CTF single”, the cross sub channel transportation induced by turbulent mixing and void drift is ignored Therefore, the comparison between results by “CTF single” and CFX is more appropriate due to the same boundary conditions Table 4.7 Void fraction prediction by CTF and CFX at downstream of channel at z = 3.48m Case ID CTF bundle CFX 0.062 CTF Boiling CTF single Mode nucb 0.094 SB01003-16-15 SB01003-16-30 0.146 nucb 0.101 0.1371 SB01003-14-34 0.173 nucb 0.152 0.1964 SB01003-20-15 0.153 nucb 0.2 0.2195 LB01002-20-18 0.424 nucb 0.356 0.2979 LB01002-15-30 0.395 nucb 0.361 0.32 LB01002-20-20 0.442 nucb 0.438 0.3954 0.1319 102 SB01003-09-37 0.429 nucb 0.444 0.4064 LB01002-30-30 0.609 nucb 0.64 0.6433 The column “CTF Boiling Mode” shows the heat transfer mode in CTF results at givem location (z=3.48 m) The two last columns in Table 4.7 shows the void fraction prediction by “CTF single” and CFX for nine cases and Figure 4.18 shows the behavior of void fraction along the sub channel between “CTF single” and CFX As conclusions given in section 3.5 chapter 3, the CTF always gives under void fraction prediction when αg