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BỘ GIÁO DỤC VÀ ĐÀO TẠO TRƯỜNG ĐẠI HỌC BÁCH KHOA HÀ NỘI HOÀNG MINH GIANG NGHIÊN CỨU HIỆN TƯỢNG CHUYỂN PHA TRONG VÙNG HOẠT LÒ PHẢN ỨNG LUẬN ÁN TIẾN SĨ CƠ HỌC Hà Nội – 2016 BỘ GIÁO DỤC VÀ ĐÀO TẠO TRƯỜNG ĐẠI HỌC BÁCH KHOA HÀ NỘI LỜI CAM ĐOAN Văn Hiền. HOÀNG MINH GIANG NGHIÊN CỨU HIỆNkết TƯỢNG CHUYỂN PHAtrình TRONG VÙNG Các số liệu, những luận nghiên cứu được bày trong luậnHOẠT văn PHẢN ỨNG này trung thực và chưa từng đượcLÒ công bố dưới bất cứ hình thức nào. Tôi xin chịu trách nhiệm về nghiên cứu của mình. GV Hướng dẫn sinh Chuyên ngành: CƠ HỌCNghiên CHẤT cứu LỎNG Mã số: 62440108 Nguyễn Đông LUẬN ÁN TIẾN SĨ CƠ HỌC NGƯỜI HƯỚNG DẪN KHOA HỌC: 1. PGS.TS NGUYỄN PHÚ KHÁNH 2. TS TRẦN CHÍ THÀNH Hà Nội – 2016 2 LỜI CAM ĐOAN Tôi xin cam đoan luận án là công trình nghiên cứu của bản thân tôi dưới sự hướng dẫn của tập thể giáo viên hướng dẫn. Các kết quả nêu trong luận án là trung thực, không sao chép của bất kỳ công trình nào và chưa từng được công bố trong bất kỳ công trình nào khác. Hà Nội, ngày 27 tháng 4 năm 2016 NGHIÊN CỨU SINH HOÀNG MINH GIANG Hướng dẫn 1 PGS. NGUYỄN PHÚ KHÁNH Hướng dẫn 2 TS. TRẦN CHÍ THÀNH 3 LỜI CẢM ƠN Trước hết, tôi xin bày tỏ lòng kính trọng và biết ơn tới: PGS Nguyễn Phú Khánh và TS Trần Chí Thành, những người thày đã trực tiếp hướng dẫn, giúp đỡ tôi trong quá trình học tập và thực hiện luận án. Tôi xin chân thành cảm ơn các thày cô tại Bộ môn Kỹ thuật Hàng không và Vũ trụ, Viện Cơ khí Động lực; cảm ơn TS Lê Văn Hồng, Viện Năng lượng Nguyên tử Việt Nam, chủ nhiệm đề tài độc lập cấp nhà nước (mã số ĐTĐL.2011-G/82) “Nghiên cứu, phân tích, đánh giá và so sánh hệ thống công nghệ nhà máy điện hạt nhân dùng lò VVER-1000 giữa các loại AES-91, AES92 và AES-2006”, các đồng nghiệp Hoàng Tân Hưng, Trung tâm An toàn hạt nhân, Nguyễn Hữu Tiệp, Trung tâm Năng lượng hạt nhân, Viện Khoa học và Kỹ thuật hạt nhân đã giúp đỡ, tạo điều kiện để tôi có thể hoàn thành luận án này. Tôi cũng xin trân trọng cảm ơn Ban lãnh đạo Viện Khoa học và Kỹ thuật hạt nhân, Viện đào tạo Sau đại học của Trường Đại học Bách Khoa Hà Nội đã cử tôi đi đào tạo cũng như tạo điều kiện thuận lợi trong quá trình thực hiện luận án. Hà nội ngày 27/4/2016 Nghiên cứu sinh Hoàng Minh Giang 4 STUDY ON PHASE CHANGE IN THE CORE OF NUCLEAR REACTOR 5 TABLE OF CONTENTS Abbreviations and Nomenclature ...............................................................................................................8 List of Tables ..............................................................................................................................................12 List of Figures .............................................................................................................................................14 Overview ....................................................................................................................................................17 Chapter 1. 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 2. 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 3. 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 6 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 4. 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 7 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 4 x 4 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 8 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 9 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 3 (source [35]). ........................................................49 Table 3.4 Chronological sequence of Event 3 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 1.........................................................................................................62 Table 3.15 y+ predicted by Mesh 2.........................................................................................................62 Table 3.16 y+ predicted by Mesh 3.........................................................................................................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 4.18 Axial sub channel void fraction prediction by CFX and CTF ..........................................103 Figure 4.19 (a) Overview of mesh (b) Zooming of mesh ....................................................................105 Figure 4.20 Four cases with specific timing for study by CFX...........................................................105 Figure 4.21 Improvement by CFX in left pictures and upper and lower bounds in right pictures .......107 16 Overview Phase change in the nuclear reactor core is related to safety criteria such as Departure of Nucleate Boiling (DNB) during normal and transient conditions. So that, a lot of computer codes with verification and validation against experiment are used to investigation of thermal hydraulics behavior of vertical boiling flow in core channel with system and component scales. Until now, even many studies on boiling flow are implemented in CFD scale codes, but their utilization to specific nuclear reactor is not yet applied. Thus, the utilization of many codes including CFD scale (Ansys CFX) to investigate void fraction in hot channel of VVER1000/V392 reactor core is studied in this work. Due to VVER-1000/V392 nuclear reactor is a candidate for Ninh Thuan 1 nuclear power project, so that the understanding of VVER’s reactor technologies including research works of this thesis is important to develop competence of nuclear safety in Vietnam. In this thesis, the numerical simulation is used to investigate boiling flow in the core channel of VVER-1000/V392 reactor with verification and validation against experiment with similar Pressurized Water Reactor conditions. The thesis includes four chapters together a conclusion in the last. Chapter 1 mentions about introduction that leads to motivation of this study. Chapter 2 presents the methodology related to multi scale analysis along with the code theories at different scale for RELAP5, CTF and Ansys CFX with focus on phase change models. The verification and assessment of modeling used in these codes versus experiment data are presented in chapter 3. The system simulation results are compared with those in SAR documents. The assessment of CTF code is implemented by simulation BM ENTEK experiment tests which is an International Standard Benchmark to investigate boiling flow through Russian fuel bundle of RBMK reactor. The meso scale code Ansys CFX is verified with PSBT single sub channel which is also an International Standard Benchmark as well. Chapter 4 presents the simulation of VVER1000/V392 by three scales with system, component and CFD codes corresponding with RELAP5, CTF and Ansys CFX, respectively. Void fraction in hot channel of the core is predicted by utilization of CTF and Ansys CFX codes. It is summarized several main contributions from the thesis as following:    It is proposed a reality of best estimate approach in void fraction prediction by utilization of multi codes and multi scale including MCNP5, RELAP5, and CTF for analysis of void fraction behavior in the core during transient. It is established a procedure of utilization of CTF and Ansys CFX for improvement of void fraction prediction as following: (a) at sub cooled region, corresponding with small bubble flow regime, Ansys CFX results is used; (b) in saturated boiling region, CTF and Ansys CFX void fraction curves along the channel is used as upper bound and lower bound to predict void fraction in the core. It is found that, in saturated boiling region, the wall boiling model built in Ansys CFX is incorrectly partitioned heat flux to corresponding parts in convective, quenching and evaporative. This issue causes Ansys CFX gives under prediction of void fraction in saturated boiling region. It is proposed a calibration for bubble departure diameter and 17 maximum area fraction to improve void fraction prediction by Ansys CFX in saturated region. 18 Chapter 1. Introduction to research work 1.1 Status of nuclear power in the World and Vietnam Nuclear technology uses the energy released by splitting the atoms of certain elements. After Second World War, nuclear technology turned to peaceful purposes of nuclear fission for power generation. Today, as updated in February 2015 [46], the world produces as much electricity from nuclear energy as it did from all sources combined in the early years of nuclear power. Civil nuclear power now can boast over 16,000 reactor years of experience and supplies almost 11.5% of global electricity needs. The 31 countries host over 435 commercial nuclear power reactors with a total installed capacity of over 375,000 MWe as illustrated in Figure 1.1. About 70 further nuclear power reactors are under construction, equivalent to 20% of existing capacity, while over 160 are firmly planned, and equivalent to half of present capacity. Figure 1.1 Nuclear power generation by country in 2013 (source [46]) After Fukushima accident in 2011, it is needed to improve performance from existing nuclear reactor safety, so that stress test is implemented for every reactor to ensure that it can stand with design extension conditions as it occurred in Fukushima. Nuclear power program in Vietnam was confirmed by National Assembly Decision on November 25, 2009 with a plan to build four nuclear power units. Russian VVER technology was selected for the first site and the VVER-1200/V491 and VVER-1000/V392 nuclear reactors were considered as the candidates. In parallel with capacity building for nuclear 19 power infrastructure, Vietnam Atomic Energy Institute (VINATOM) has been assigned to develop technical competence of nuclear safety and, in future, it will become a main Technical Support Organization (TSO) for nuclear safety in Vietnam. 1.2 Brief overview of nuclear safety In any nuclear reactor, nuclear safety includes three primitive principals related to all operational conditions: (a) control of reactivity, (b) heat removal from the core to ultimate heat sink and (c) confinement of fission products in case of accident occurrence. Requirements of nuclear safety enforce the utilization of defend in depth (DID) policy in nuclear power plant design. In general, defense in depth policy is implemented by multiple levels of protection (Table 1.1) and protection barriers Table 1.1 Multiple levels of protection from DID approach (source [45]) Levels Level 1 Objective Prevention of abnormal operation and failures Level 2 Control of abnormal operation and detection of failures Level 3 Control of accidents within the design basis Control of severe plant conditions, including prevention of accident progression and mitigation of the consequences of severe accidents Mitigation of radiological consequences of significant release of radioactive materials Level 4 Level 5 Essential Mean Conservative design and high quality in construction and operation Control, limit & protection systems and other surveillance feature Engineered safety features and accident procedures Complementary measures and accident managements Off-site emergency response For PWR reactors at power operation, the barriers confining the fission products are typically: (a) fuel matrix, (b) fuel cladding, (c) boundary of the reactor coolant system, (d) containment system as shown in Figure 1.2. The nuclear utility owner must provide and demonstrate that their technical plan design is satisfied for the safety requirements through safety analysis report (SAR) which is reviewed and approved by nuclear regulatory authorities (NRA) and independent TSO. All main issues, related to nuclear safety during plant life time, are mentioned in SAR as illustrated in Table 1.2. Nuclear safety covers a wide range of issues related to the plant including external hazards such as seismic, tsunami, flooding… and internal hazard resulted from failure of structures, systems and components together with human factor that are important to safety. In chapter 15 of SAR, thermal hydraulics safety analysis is performed by simulation of different categories of postulated transient and design base accident such as reactivity insertion accident (RIAs), loss of coolant flow (LOFAs) and loss of coolant accident (LOCAs). The main system and other connected systems including reactor coolant system (primary system), secondary system are modeled by system code that 20
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