Nghiên cứu hiện tượng chuyển pha trong vùng hoạt lò phản ứng

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 NGHIÊNliệu, kết luận nghiên cứu đƣợc trình bày luận văn Các số CỨU HIỆN TƢỢNG CHUYỂN PHA TRONG VÙNG HOẠT LÒ PHẢN ỨNG trung thực chƣa đƣợc công bố dƣới hình thức Tôi xin chịu trách nhiệm nghiên cứu GV Hướng dẫn Chuyên ngành: CƠ HỌCNghiên cứu sinh CHẤT LỎNG Mã số: 62440108 Nguyễn Đông LUẬN ÁN TIẾN SĨ CƠ HỌC NGƢỜI HƢỚNG DẪN KHOA HỌC: PGS.TS NGUYỄN PHÚ KHÁNH TS TRẦN CHÍ THÀNH Hà Nội – 2016 LỜI CAM ĐOAN Tôi xin cam đoan luận án công trình nghiên cứu thân dƣới hƣớng dẫn tập thể giáo viên hƣớng dẫn Các kết nêu luận án trung thực, không chép công trình chƣa đƣợc công bố công trình khác Hà Nội, ngày 27 tháng năm 2016 NGHIÊN CỨU SINH HOÀNG MINH GIANG Hƣớng dẫn PGS NGUYỄN PHÚ KHÁNH Hƣớng dẫn TS TRẦN CHÍ THÀNH LỜI CẢM ƠN Trƣớc hết, xin bày tỏ lòng kính trọng biết ơn tới: PGS Nguyễn Phú Khánh TS Trần Chí Thành, ngƣời thày trực tiếp hƣớng dẫn, giúp đỡ trình học tập thực luận án Tôi xin chân thành cảm ơn thày cô Bộ môn Kỹ thuật Hàng không Vũ trụ, Viện Cơ khí Động lực; cảm ơn TS Lê Văn Hồng, Viện Năng lƣợng Nguyên tử Việt Nam, chủ nhiệm đề tài độc lập cấp nhà nƣớc (mã số ĐTĐL.2011-G/82) ―Nghiên cứu, phân tích, đánh giá so sánh hệ thống công nghệ nhà máy điện hạt nhân dùng lò VVER-1000 loại AES-91, AES92 AES-2006‖, đồng nghiệp Hoàng Tân Hƣng, Trung tâm An toàn hạt nhân, Nguyễn Hữu Tiệp, Trung tâm Năng lƣợng hạt nhân, Viện Khoa học Kỹ thuật hạt nhân giúp đỡ, tạo điều kiện để hoàn thành luận án Tôi xin trân trọng cảm ơn Ban lãnh đạo Viện Khoa học Kỹ thuật hạt nhân, Viện đào tạo Sau đại học Trƣờng Đại học Bách Khoa Hà Nội cử đào tạo nhƣ tạo điều kiện thuận lợi trình thực luận án Hà nội ngày 27/4/2016 Nghiên cứu sinh Hoàng Minh Giang STUDY ON PHASE CHANGE IN THE CORE OF NUCLEAR REACTOR 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 ̅ h h h h h hc hl* h hg h h ̇ ̇ 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 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