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 Hà Nội - 2016 9IO£-Ị0N?H OỎH 03 IS NẸLL MỴ Nỹm 8010171739 :ọs ĐMỌ1XVH3 3ỎH 03 •qu?ẵn usẤniỊO DN.n NỴHd Ọ1 XVOH DNJXA ĐNOMX VHd NẸAÍ1XD ONQ.nX NỆIH OỢ3 MẸĨHOM OMVIO IIVIIV OMVOH IỘM VH VOI IM II3YÍI 30II IV CI 3YỌ.1MI OVX OVQ VA 3íìa OVIĐ Oe LỜI CAM ĐOAN Tôi xin cam đoan luận án công trình nghiên cứu thân hướng dẫn tập thể giáo viên hướng dẫn Các kết nêu luận án trung thực, không chép công trình chưa công bố công trình khác Hà Nội, ngày 27 tháng năm 2016 NGHIÊN CỨU SINH HOÀNG MINH GIANG Hướng dẫn Hướng dẫn PGS NGUYỄN PHÚ KHÁNH TS TRẦN CHÍ THÀNH LỜI CẢM ƠN Trước hết, xin bày tỏ lòng kính trọng biết ơn tới: PGS Nguyễn Phú Khánh TS Trần Chí Thành, người thày trực tiếp hướng 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 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 so sánh hệ thống công nghệ nhà máy điện hạt nhân dùng lò VVER-1000 loại AES-91, AES- 92 AES-2006”, đồng nghiệp Hoàng Tân Hưng, Trung tâm An toàn hạt nhân, Nguyễn Hữu Tiệp, Trung tâm Năng lượng 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 Trường Đại học Bách Khoa Hà Nội cử đào tạo tạo điều kiện thuận lợi trình thực luận án Hà nội ngày 27/4/2016 Nghiên cứu sinh STUDY ON PHASE CHANGE IN THE CORE OF NUCLEAR REACTOR Hoàng Minh Giang 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 COBRATF 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 acrit 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 a ^mtscv A1" ■rtint,shl A'” , rMnt,shv As Ax Cpl Cpv G h 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) g,sat int,scl h int,scv h int,shl h int,shv hc hl* Vapor saturation enthalpy (J/kg) Sub-cooled liquid interface heat transfer coefficient (W/m K) vapor interface heat transfer coefficient Sub-cooled (W/m K) Super-heated liquid interface heat transfer coefficient (W/m K) Super-heated vapor interface heat transfer coefficient (W/m2.K) Chen correlation heat transfer coefficient (W/m2.K) Liquid enthalpy (J/kg) h l,sat g h / h ig mr m™ m Plk Liquid saturation enthalpy (J/kg) Vapor enthalpy (J/kg) Vapor interface heat transfer coefficient (W/m3.K) Q 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) h h wf Qconv QW ift Qboil T T gS r-J^ T crit Tl, Tf rb «ki a kiEQ PTP Pki Pl Pv P' P m pî 5 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) 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) r hgw Vapor generation from near wall (kg/m3.s) Total Vapor Generation (kg/m3.s) AX a li P Mesh-cell axial height (m) Surface tension (N/m) Fluid viscosity (Pa.s) Pressure (Pa) r’’ Tw Evaporation rate (kg/m2.s) Wall surface temperature (K) Tchf ,Tcrit Re Critical heat flux temperature (K) Reynolds number Pr Nu n k , l ^i hv hnb 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/m Liquid.K) enthalpy (J/kg) h l g h fc h f h c h g FChen f Dh Cp Ax As 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) Fo 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 Pa P a ,r aa Rw Rw Rq Re' Rc dp 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) AX XTT PI T sat Ti sk Schen Qi G ^2 Chen suppression factor Heat transfer per volumetric unit (W/m3) Mixing mass flux (kg/m3.s) Area influence factors all three cases as presented in Table 4.6 Therefore, the mesh Ml is used to investigate the remaining cases in Table 4.5 Table 4.6 Average void fraction for different meshes Case ID SB01003-09-37 SB01003-14-34 SB01003-20-15 Void of mesh Ml at 3.48m 0.4064 0.1964 0.2195 Void of mesh M2 at 3.48m 0.4093 0.1960 0.2185 Void of mesh M3 at 3.48m 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 CTF Boiling CTF single Mode SB01003-16-15 nucb 0.094 0.1319 0.062 SB01003-16-30 0.146 nucb 0.1371 0.101 SB01003-14-34 0.173 nucb 0.152 0.1964 SB01003-20-15 0.153 nucb 0.2195 0.2 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 SB01003-09-37 LB01002-30-30 0.429 0.609 nucb nucb 0.444 0.64 0.4064 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 a g 100Ỉ-07-CTF-MI I3otoo:-07-Cfy-Ni J _ ■ _ 4.7 Conclusions In summary of the work done in this chapter, several main issues from study of void fraction prediction for hot channel in the core of VVER-1000/V392 are presented as below: ■ It is well implemented the simulation of VVER-V392 reactor by system code RELAP5 with power distribution provided by neutron code MNCP5 for the first fuel cycle using the model which is validated in chapter ■ Due to the fact that RELAP5 is 1D code then core modeling flow is described as a pipe with all cross section area belonging to a control volume It leads to define the same value of liquid temperature in cross section area Then, if simulate a bundle of channel in core with hydro equivalent diameter rather large (more than several cm), it leads to the fact that temperature of liquid near wall is not determined with enough accuracy for phase change models and in this situation, RELAP5 is not appropriate tool to predict void fraction ■ CTF is accepted tool to predict void fraction in core for both of sub channel or bundle of sub channels However, due to a huge number of sub channels and gaps in modeling of a fuel bundle, only a part of fuel assembly is practical to put in modeling ■ CTF and CFX can be used to predict void fraction in the core based on reference to each other With the void fraction below 0.2 and heat transfer mode in CTF is sub cooled boing, CTF tends to give under prediction and in this case CFX gives better prediction with accuracy around ±0.03 of void ■ In saturated boiling region, the wall boiling model RPI built in CFX is not partitioned correctly heat fluxes to corresponding parts of convective, quenching and evaporative Therefore, more heat is transferred to convective part and it makes liquid temperature near heated wall higher than saturated one This issue shows that CFX gives under prediction of void fraction in saturated boiling region Because CTF tends to give over prediction of void fraction in nucleate boiling mode So that, in saturated boiling region, CTF and CFX void fraction predictions can be used as upper bound and lower bound curves Conclusions and proposals Achievements and new findings given by the thesis As mentioned in chapter 1, the thesis objective focuses on void fraction prediction in the core of VVER-1000/V392 reactor with the goal as following: ■ To adopt a procedure of void fraction prediction during transient using multi scale analysis based on the computer codes: MCNP5, RELAP5 and CTF; ■ To consider a combination of CTF and Ansys CFX codes to improve void fraction predicted by CTF in specific timing within the transient period In order to establish the appropriate simulation modeling given by the thesis, the verification and validation of simulation by comparison calculation results versus the experiment data or results from safety analysis report is implemented carefully The role of experiment data is important to validate physical models of computer codes to make sure if they are appropriate with specific conditions of the problems Therefore, a lot of technical works are done in the thesis such as development of RELAP5 input desk for the VVER-1000/V392 including simulation of normal and safety systems The simulation modeling of VVER-1000/V392 reactor by system code RELAP5 is compared with those in Belene safety analysis report in steady state and LOCAs accidents as mentioned in section 3.2 of chapter Based on this modeling, the autonomy of VVER-1000/V392 in the design extension condition with specific scenario of Large Break LOCA and SBO simultaneous occurrence is investigated in the author’s paper (numbering [1] in section “List of Author’ papers and report”) which contributes to understanding of VVER’s passive safety system capability The validation of CTF boiling models in high pressure conditions such as BM ENTEK and PSBT experiments exposes that the correlations of wall heat transfer to water as well as evaporation and condensation terms at the interfacial area are needed to be considered due to the fact that CTF tends to under predict void fraction in sub cooled region where void fraction below 0.2 and tend to over predict void fraction at nucleate boiling region where void fraction above 0.2 The above investigation and conclusions can be found in section 3.3 and 3.4 of chapter as well as mentioned more detail in Author’s papers (numbering [4,5] in section “List of Author’ papers and report”) The verification and validation of Ansys CFX (shortly written CFX) code for two phase flow in PWR conditions is implemented by simulation of the runs for single sub channel in PSBT benchmark This works is really a challenge in all aspects such as setting up of appropriate models, getting suitable convergence criteria and finally accuracy of simulation results in comparisons with measured values in experiment Regarding to the convergence of S1 runs 1.4325, the criteria of RMS of x 10 -5 cannot reached and is reported in [24] (J Weis et al report at the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics NURETH ’14) As shown in section 3.4.3 of chapter and also mentioned in the Author’s paper (numbering [2] in section “List of Author’ papers and report”), this issue is solved by Author’ proposal of a new setting up for two phase flow as presented in section 3.4.3 or by utilization of Multi Size Group sub model (MUSIG) for the different bubble diameters in the bulk of water The stability of setting up model for two phase flow given by the thesis is mentioned in section 3.4.5 of chapter through the sensitivity study on physical models The works at this section show the specific role of each none drag force term such as lift force, wall lubrication force and turbulent distribution force coefficient It is concluded that the closure models related to momentum equations, in general, affect to void fraction in cross section of channel but not amount of vapor For the study on evaporation and condensation models in CFD level, it is shown that the most sensitive factor to evaporation is bubble departure diameter and the sensitivity of Nusselt number correlation is also significant affected to condensation The sensitivity of void fraction prediction caused by change of bubble departure diameter is also presented in Author’s report (numbering [3] in section “List of Author’ papers and report”) Several interesting issues are found on section 3.4.6 of chapter when study on assessment of CFX and CTF modeling results in comparison with PSBT single channel It is found that CFX with given setting up models give better void fraction prediction in comparison with experiment while CTF tends to give under prediction in sub cooled region with void fraction below 0.2 In the saturated region corresponding with small-to-large bubble flow regime, it is found that CTF give the over prediction in this region while CFX give the under prediction The reason that makes CFX giving under void fraction prediction in the saturated region results from wrong heat partition of wall boiling model RPI employed in CFX Heat transfer to liquid as convective and quenching parts is too high and lead to limit heat transfer to evaporative part due to total heat flux from wall to boiling flow is fixed as discussed in section 3.4.7 of chapter A new achievement of study on void fraction prediction by CFX simulation is presented in section 3.4.8 where, at the Author’s point, two factors: bubble departure diameter and maximum area fraction MAF are needed to be scaling up in order to reduce the difference between average temperature of liquid along radial distribution and saturated point It is expected that a new paper based on this study will be published in future As mentioned in chapter 4, a procedure to conduct multi scale analysis of thermal hydraulics for VVER-1000/V392 is introduced The results from code MCNP5 provide with axial channel power distribution, hot channel peaking factor for simulation by system code RELAP5 and also provide with axial channel power distribution, relative power distribution in hot channel bundle for simulation by CTF The output results from code RELAP5 provide initial and boundary conditions such as pressure at inlet, power, mass flow rate at inlet and inlet coolant temperature for sub channel level simulation by CTF in both steady state and transient conditions With the purpose of combination of CTF and CFX codes to improve void fraction predicted by CTF in specific timing within the transient period, the initial and boundary condition used to establish simulation models in CFX are provided by MCNP5 with power distribution and by CTF with thermal hydraulics conditions It is concluded that void fraction prediction in core is improved by the utilization of CTF and CFX codes with reference to each other CTF give the void prediction during transient time and its results in specific moment is improved by reference with CFX In the sub cooled region with low void (void below 0.2 defined by CTF) the CFX result is used and in the saturated boiling region (defined by CTF with void normally between 0.2 and 0.5) then void fraction prediction curves calculated by CTF and CFX is considered as upper and lower bounds To finalize the thesis results, it is summarized the main findings and achievements presented in chapters and as following: ■ It is proposed a reality of best estimate approach in void fraction prediction by utilizing multi code and multi scale including MCNP5, RELAP5, CTF and CFX for analysis of void fraction behavior in the core during transient ■ For system analysis by RELAP5 code for VVER-1000/V392, it is found that temperature near heated wall is not defined with enough accuracy due to large equivalent diameter if simulation a whole of fuel assembly, so the phase change models in RELAP5 not give appropriate value of void fraction ■ From verification and validation of CTF results with ENTEK BM experiment, it is observed that CTF tends to give under prediction of void in the region of sub cooled boiling and flow regime in small bubble (a g < 0.2) and CTF tends to give over prediction of void in nucleate boiling region, corresponding with small-to-large bubble in flow regime ■ From verification with PSBT single sub channel experiment, CFX with model setup proposed in this thesis is converged with RMS of 1e-6 and stabilized in term of average void fraction prediction with physical sensitivity study For the sub cooled boiling region corresponding with small bubble of flow regime (a g < 0.2), CFX gives the appropriate void fraction prediction with accuracy around ±0.03 of void ■ In saturated boiling region, the wall boiling model built in CFX is incorrectly partitioned heat flux to corresponding parts in convective, quenching and evaporative This issue causes CFX gives under prediction of void fraction in saturated boiling region ■ It is proposed a calibration for bubble departure diameter and maximum area fraction to improve void fraction prediction by CFX in saturated region ■ It is established a procedure of utilizing CTF and CFX codes for void fraction prediction as following: (a) at sub cooled region, corresponding with small bubble flow regime, CFX results is used; (b) in saturated boiling region, CTF and CFX void fraction curves along the channel is used as upper and lower bound to predict void fraction in the core Proposal of future work Utilization of CFD codes for investigation of void fraction in the core is still a challenge This comes from complexity of boiling phenomena and the lack of experiment with similar PWR condition to verification and validation CFD models Based on study in the thesis, several following issues are proposed to study ■ Study on modification of RPI wall boiling model built in CFX (and FLUENT) in saturated boiling region Due to the fact that, in saturated boiling model, liquid temperature is the same saturated one everywhere, even near wall, so that only evaporation and quenching phenomena can occur 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Atomic Power Laboratory, 1970 [41] S.C.P Cheung, S Vahaji , G.H Yeoh , J.Y Tu (2014) Modeling sub cooledflow boiling in vertical channels at low pressures - Part 1: Assessment of empirical correlations Article in press, International Journal of Heat and Mass Transfer xxx (2014) xxx-xxx [42] S.C.P Cheung, S Vahaji , G.H Yeoh , J.Y Tu, (2014) Modeling sub cooledflow boiling in vertical channels at low pressures - Part 2: Evaluation of mechanistic approach Article in press, International Journal of Heat and Mass Transfer xxx (2014) xxx-xxx [43] Th Frank, F Reiterer and C Lifante (2011) Investigation of the PWR Sub channel Void Distribution Benchmark (OECD/NRC PSBT benchmark) using Ansys CFX Proceedings of the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH ’14) , Hilton Toronto Hotel, Toronto, Ontario, Canada, September 25-29, 2011 [44] Vinay Karanam (2011) Evaluation of sub-channel flow mixing coefficient for typical PWR fuel bundles having spacers using CFD analysis A master thesis at Homi Bhabha National Institute, 2011 Pp 17-19 [45] Won-Pil Baek (2009) Overview of PWR Safety Analysis Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute VAEI Training Course on Fundamental Safety Analysis and Computer Code with Basic T/H Knowledge Requirements, October 12-16, 2009, Hanoi, Vietnam [46] World Nuclear Association, “http://www.world-nuclear.org” List of Author’ papers and report [1] Hoang Minh Giang, Nguyen Phu Khanh, Le Van Hong., (2014) “Capability analysis of passive systems in typical design extension conditions for nuclear reactor VVER1000/V392”, Journal of Science and Technology 52 (2C) (2014) pp 81-92 [2] Hoang Minh Giang, Nguyen Phu Khanh, Le Van Hong, Le Dai Dien.,(2014) “Study on improvement of convergence for PWR sub channel void distribution benchmark”, Journal of Science and Technology 52 (2C) (2014) pp 184-197 [3] Hoang Minh Giang, Nguyen Phu Khanh., (2014) “Numerical investigation of departure bubble diameter for wall boiling model in PWR sub channel”, proceeding in AUN/SEED-Net Regional Conference on Mechanical and Manufacturing Engineering, 9-10 October, 2014 (RCMME-2014) [4] Hoang Minh Giang, Hoang Tan Hung, Nguyen Phu Khanh., (2015)“Investigation of CTF void fraction prediction by ENTEK BM experiment data”, Nuclear Science and Technology (ISSN 1810- 5408),vol5, No1, 2015 pp.8 -17 [5] Hoang Minh Giang, Hoang Tan Hung, Nguyen Huu Tiep., (2015) “Multi codes and multi-scale analysis for void fraction prediction in hot channel for VVER-1000/V392”, Nuclear Science and Technology (ISSN 1810-5408), Vol5, No3, 2015 pp.19-31 [6] Hoang Minh Giang, Hoang Tan Hung, Nguyen Phu Khanh., (2015) “Numerical study on CTF code to predict void fraction in PWR sub channel conditions”, Nuclear Science and Technology (ISSN 1810-5408),Vol5, No4, 2015 pp.30-38 [...]... integration, or space filtering And for two-phase flow only: 4 Choice of the number of phases or fields of the model by multiplying basic equations by phase characteristic functions or field characteristic functions 5 Treatment of interface, which can be Deterministic Interface (DI), Filtered Interface (FI) or Statistical Interface (SI) It is observed that, for single phase flow, only three main types of... simulating even horizontal two phase flow which is not available in the version before The RELAP-3D and MARS-3D are resulted from coupling RELAP5 with COBRA-TF with purpose of better simulation core and steam generator in nuclear power plant Recently, an extension of CFD code application for two-phase flow is implemented as a part of multi-scale of thermal hydraulic safety analysis Two-phase flow CFD used for... implied ■ ■MMf MM Seide« or tverag fateflnf ini Jj} fixàvct MPPI N h Huid/solid rharacî»rmKj Figure 2.6 The tree of two-phase thermal hydraulic modeling approaches (source [11]) 2.2 Phase change models in system code RELAP5 The RELAP5 models include conservation equations for two-phase flow with one dimensional simulation of thermal hydraulics behavior along flow path The detail of RELAP5 conservation... transfer Thus, the phase change models are related to interface energy and mass exchange and wall heat transfer The vapor generation or condensation [19] in RELAP5 is modeled by phase change induced from heat transfer between interface and wall heat transfer effect: H i„(Ts-T„) + H i f ( T s — T f ) , ^ (2.1) [Hg ^ The item, [h'g, h|), is chosen similar to ( h g , h g ) but it denotes for phasic enthalpies... so that study on void fraction in transient condition is the first step to approach understanding DNB mechanism 1.3.1 Role of void fraction in simulation of two phase flow The value of void fraction plays an important role in modeling of two phase flows During solving the conservation equations, the void fraction is calculated Then, the flow regime is defined based on the value of void fraction For... on Figure 2.6, it is shown that RELAP5 code belongs to column (1D nF) with one dimension simulation and two phase conservation equations Space treatment is applied to many quantitative terms such as temperature, pressure and so on Time treatment is applied to some terms such as velocity and phase characteristic is identified Similarly, CTF code falls in column “Porous-3D, nF) in Table 6 with requirement... during 40 seconds of transient condition at the beginning of LOCAs with different break sizes 1.5.2 Scope of study It is also limited the scope of the study due to complexity of the two phase flow The investigated two phase flow through core sub channels is vertical flow with the specific regime such as bubbly, slug, churn and annular The left picture and the right of Figure 1.7 shows the temperature... [11]) CFD in open System code Component code medium RANS 3D & Type of model 0D, 1D, 2D and porous 3D & porous 3D sub-channel analysis 3 DNS & pseudoDNS LES type No model in single-phase 106to 108 106to 108 Mixing problems in 1phase flow: boron dilution, MSLB, PTS, thermal fatigue, thermal stripping, PTS, CHF in Several days to several weeks on massively parallel compute Basic flow processes: turbulent... vapor, respectively Discussion of Increasing v phase changem near wall is 0 mentioned in [20] based0 on models and correlations, for example, the near wall evaporation is calculated by following correlation: 4 K' t “BS a DE “SA “AM 1-0 Increasing ag (2.6) Where V is the cell volume and the term Mul is given by: More detail about formula (2.7) is given in [20] 2.3 Phase change models in sub channel code CTF... heat flux and DNB The boiling crisis in flowing coolant is more complicated than pool boiling due to added effects of forced convection and bubbles clouding that tend to cover heating surface Single-phase convection 1 0 Nucleate boiling boiling film Partial boiling Film T.-T, rc] Figure 1.3 Heal flux versus temperature difference for pool boiling heat transfers (source [31]) The boiling crisis in