DSpace at VNU: Thermal hydraulic system of a VVER-1000 nuclear reactor and numerical simulations tài liệu, giáo án, bài...
VNU Journal o f Science Mathematics - Physics 27 (201 1) 111 122 Thermal hydraulic system of a VVER-1000 nuclear reactor and numerical simulations D uong Ngoc N guyen Tat Thang*’’’^ Dang Thi H oa’ ^Institute o f Mechanics, Vietnam Academy o f Science and Technology (VAST), Hanoi, Vietnam '^VNLJ University o f Engineering and Technology, Hanoi, Vietnam ^ Tokyo Institute o f Technology (TỈTECH), Tokyo, Japan Received 20 M ay 2011 A b s tra c t This paper presents som e results o f our study on the numerical sim ulation o f the them ial hydraulic system o f the Russian V V ER -1000 pressurized w ater nuclear reactors The simulations were conducted using the integrated VISA (Visual System A nalyzer) and RELAP5 (Reactor Excursion and Leak A nalysis Program , version 5) softwares known as VISA _RELA P5 Originally RELAP5 (a therm al hydraulic system code) and then recently VISA (a graphical user interface) w ere developed for the sim ulation o f the therm al hydraulic system o f W estern type pressurized w ater reactors (PW R ) undergoing transients In Vietnam , research on the num erical sim ulation o f the therm al hydraulic system o f Da Lat nuclear research reactor and some typical types o f PWRs using R ELA P5 have long been carried out in our research group along with some o f other research institutions It should be noted that know ledge o f the VVER reactor system is still lacking in Vietnam until now The data that we used in our modelings, simulations and calculations are real data o f the K alinin nuclear pow er plant (NPP) in Russia Therefore this research has important practical im plications especially for the preparation for the safe operation and proper m anagem ent o f incidents (accidents) in the N PP that will be built in N inh Thuan, Vietnam The reactors adopted in the Ninh Thuan NPP will be the Russian VVER PWR From the point of view of the available information about VVER reactors in Vietnam, our study is immensely useful since Russia has not yet much opened up infomiation about W E R reactors At this point, our research is a basic im portant step tow ards a practical study case Keywords: Therm al hydraulic, VISA , RELAP5, N uclear Reactor, Pressurized W ater Reactor (PW R ), N uclear Pow er Plant (N PP), V V E R -1000, Ninh Thuan NPP Introduction A lthough nuclear pow er industry has experienced som e'serious accidents in the past (Chernobyl accident in U kraine and Three M ile Island accident in the US) and recently the Fukushim a accident in C o rresponding author: Tel.: (+84) 983384692 Em ail: ntthang@ im ech.ac.vn 11.1 112 D.N H et al / VNU Journal o f Science, M athem atics - Physics 27 (2 Ì) 111-122 Japan, m ost people believe that nuclear pow er is still a very im portant pow er source and plays crucial role in our globe It is clearly stated that “Though nuclear pow er industry faces enorm ous safety challenges, it is still an im portant choice in the r ‘ century.” [1], Vietnam G overnm ent has approved plans to build N PPs in V ietnam that are the and the 2"*' Ninh Thuan N PPs [2], It is therefore mandatory to develop human resources for nuclear industry (especially nuclear pow er industry), and to prom ote research on various aspects related to nuclear reactors, nuclear pow er plants, nuclear safety etc A m ong those, the research and safety analysis based on numerical m odelings, calculations and sim ulations o f the reactor therm al hydraulic system using com puter codes is much im portant as well The best-estim ate thermal hydraulic simulation program s (e.g RELAP5 code [3]) have long been developed They have been step by step im proved to model more accurately the therm al hydraulic system of the nuclear reactor and NPPs Those program s are also im portant in the calculation and simulation o f im portant therm al hydraulic phenom enon in NPPs H ow ever the application o f those programs tends to be limited am ong small groups o f experts The rapid developm ent o f the capability o f the personal com puter perm its those program s now to be able to ran well on personal com puters As a consequence, those sim ulations program are becom ing more and more popular H owever, a m ajor restriction that still exists is that the preparation for the input data files is usually very com plicated and easy to have errors Therefore VISA program was developed under the cooperation betw een KAERI (Korea Atomic Energy Research Institute) and K HN P (K orea H ydro-N uclear Power), Korea to perform the tasks o f a GUI (Graphical U ser Interface) and to help users to exploit m ore effectively the thermal hydraulic sim ulation program s [4] VISA can be integrated with three therm al hydraulic simulation program s including M ARS, RETRAN-3D and RELAP5 Here we use the integrated program VISA and RELAP5 which is called V ISA _RELAP5 for short [5] VISA program has m any powerful functions to support the users in the m odeling, calculation and sim ulation o f the therm al hydraulic system , and safety assessm ent o f NPPs [6] In the world now, the RELAP5 code (developed in the US) is superior to other therm al hydraulic simulation codes in the nuclear industry and nuclear research Therefore in this research, we chose the V1SA_RELAP5 H owever the m ost im portant and crucial task is the m astering o f RELAP5 program That has long been conducted in the D epartm ent for Industrial and Environm ent Fluid D ynam ics, Institute o f M echanics, Vietnam A cadem y o f Science and Technology (V AST) through the num erical modelings, calculations and sim ulations o f Da Lat nuclear research reactor in V ietnam and som e typical PWRs using RELAP5 [7-14] The application VISA _RELAP5 to NPPs with Russian V V ER reactors is one o f the following steps to contribute to raising our capability in thermal hydraulic research and safety analysis, and to mastering the technology o f V VER reactors that will be transferee! to V ietnam in near future Prelim inary results o f this research have been reported at the IX N ational C onference on N uclear Science and T echnology held in A ugust 2011 in Ninh Thuan province, V ietnam [15], Through our com m unication w ith other nuclear research groups com ing to the conference from alm ost all o f nuclear research institutions in Vietnam and froữi abroad, our study would be the first o f this kind in Vietnam G iven the fact that know ledge o f Russian VVER reactor system is still seriously lacking in V ietnam (perhaps to some extent, even in the world [16]), our research hopefully will provide prelim inary relatively detailed inform ation about W E R reactors This type o f reactors has som e characteristics different from the W estern type PW R reactors V V ER reactors have horizontal steam generators (rather than vertical ones in PW Rs), circulation cooling loops having isolation valves that can be closed to isolate one (or several) loop(s) if necessary (e.g when the reactor operates at low pow er level or in case o f em ergency etc.) [17, 18], D.N H et a l / VNU Journal o f Science M athem atics - Physics 27 (2011) 111-122 113 Brief of VISA and RELAP5 codes 2.1 VISA The V ISA graphical user interface program was originally developed under the jo in t effort o f some research institutions and energy industries in South K orea [4-6], This program is designed to be integrated with som e therm al hydraulic sim ulation program s such as RELA P5, M ARS and R ET R A N etc through the use o f new w ritten or m odified inpuưoutput functions (not calculation-related functions) o f the therm al hydraulic sim ulation programs So the connection betw een V IS A interface and the sim ulation program s is by the help o f only input/output functions (the calculation-related functions o f the sim ulation program s are kept intact) The connection is via the dynam ic link libraries (D LLs) o f the sim ulation program s [4-5] M ain functions o f V ISA include: • • • • • M anage inpuưoutput files, the graphic files; select unit (SI or B ritish etc.) for the output results (Project functions); View, edit and change the value o f the param eters o f the control system, the geom efrical param eters o f therm al hydraulic system (Pre-processor functions); G raphically view the calculated results directly during the calculation process and review the previously calculated results by using graphics and mimics (G raphic and M im ic functions); View the output results through graphical windows, graphs o f variables (norm ally the variables o f time); m onitor the status o f trips (G raphic interface and Trip functions) Sim ulate actual operations o f the plant operators; the control o f therm al hydraulic system o f the reactor and the N PP is carried out via four types o f controls including confrols o f trips (on/off), confrols o f valve area, confrols o f flow-rate in pipes, and controls o f reactor pow er (Interactive control functions) mm sm awf o o Cl ■ It* ^ liầ i —^ ^,/V ^0 ^ Fig Graphical simulation of a typical Western PWR (Mữnic iuactions) (figure from [5]) 114 D.N H et a ỉ / V N V Journal o f Science, M athem atics - Physics 27 (2 Ỉ) Ỉ I Ỉ- 122 Ú Cắ j Ị f f ’ it« « M ■> '-ii r> »«i itm x IIAm UmM Ì I'lMdttMM*! VOID * 140 Fig G raphical presentation o f the PW R nodalization model and the calculated void in the system during calculation process (Pre-processor and Graphic functions) (figure from [5]) 2.2 RELAP5 T heR ELA PS IS a best-estim ate transient simulation code for the sim ulation o f light w ater reactor coolant system during postTilated accidents Coupled behavior o f the reactor coolant system and the reactor nineties is im plem ented RELAP5 model includes separate m odels for all o f the com ponents o f the reactor thermal hydraulic system (i.e fuel rods, reactor core, control rods, reactor vessel, pump, heat conduction structures, pipe, valves, control systems etc.) O riginally the code was based on a homogeneous equilibrium model (H EM ) o f the tw o-phase flow process ITien the code was totally rew ntten w ith the use o f a tw o-fluid, nonequilibrium, nonhom ogeneous, hydrodynam ic model for transient sim ulation o f the tw o-phase system behavior The version used in this research is RELAP5/M OD2 which em ploys a full nonequilibrium , six-equation, tw o-fluid m odel Stucy and applications o f RELAP5/M O D3.2 code was conducted in m any o f our previous researches and will not be shown here Details o f the system o f equations, the solution methods, the prograư flow chart, test calculations, and applications etc can be found in [3, 7-15, 17], Modeling, calculation and simulation of the thermal hydraulic system o f a VVER-1000 reactors using VISA RELAPS VVER reactors are o f the PW R design developed in R ussia w hich exhibit m any sim ilarities to Western PW Rs R ussia has made a lot o f effort to im prove m any aspects o f this reacto r type and to promote export to the international market One o f the potential m arket is V ietnam w hose governm ent has signed confracts with R ussian com panies to build a nuclear pow er p lan t in N in h Thuan province (the r ‘ N PP o f V ietaam ) Therefore the reactor technology, num erical m odeling, calculation, simulation and safety analysis o f the thermal hydraulic system o f V V E R nuclear pow er plant are urgently needed in V ietaam 3.Ỉ Structure o f the VVER reactor system a Overview VVER reactor system studied in this research is a VVER-lOOO/V-338 reactor (version V-338, the original cesign of the VVER-1000 nuclear reactor series with 1000 M W electric pow er) Fig below shows D.N H et a! / VNU Journal o f Science M athematics - Physics 27 (2011) I I I - I 2 115 the ty[:)ical therm al hydraulic system o f the V VER-1000 reactor C om pared with W estern PW R (Fig 1) ones can see d e a rly som e differences VVER reactors are equipped with isolation valves (Mam Gate Valve - M G V ) and horizontal steam generator (Fig 2) Table below shows som e o f the mam param eters o f the VVER-lOOOA^-338 reactor [18], In general, the V V ER nuclear reactor type has some advantages over W estern PW Rs D etailed discussions can be referred further to in [19], I'ab le M ain param eters o f the VVER-lOOOA^-338 reactor T herm al / Electric pow er C oolant Pressure (in the Prim ary system) N um ber o f cooling loops C oolant flow rate through the reactor core C oolant tem perature inlet (to reactor core) C oolant tem perature outlet (from reactor core) 3000 M W th / 1000 MW e 15.7 M Pa 84800 mVh 289.7 “C 320.0°c Fig O verview o f the therm al hydraulic system o f a typical VVER reactor (horizontal steam generator ) The main com ponents o f the therm al hydraulic system o f the V V ER reactor including (Fig 3): - Reactor vessel - Reactor core - Control rods - P re s s u riz e r 5, - H ot and cold legs (prim ary system) 7, - M a m G a te V a lv e s - H orizontal Steam G enerator 10 - M am circulation pum p (prim ary system) 11 - Steam line 12 - T u rb in e s 13 - C old leg (secondary cooling system) 14 - Control rod driving m echanism s b R eactor vessel The design o f the reactor vessel and the vessel’s main dim ensions are shown in Fig Mam param eters are show n in T able below 116 D.N Hai et a ỉ / VNU Journal o f Science, M athem atics - Physics 27 (20Ĩ Ỉ) Ỉ Ỉ Ỉ - Ỉ 2 ‘ I1 I ; ! r r m ầ _ .l y N t J / Ị : Fig VVER-lOOOA^-338 reactor vessel Table M ain param eters o f the VV£R-1000A ^-338 reactor vessel Length (m) Diam eter, external on the cylindrical part o f the reactor (m in) Thickness o f the cladding layer (mm) nozzles, diam eter (m m ) R eactor vessel w eight (ton) R eactor vessel total volum e (m^) Core volum e (m^) Core heat transfer surface (m^) Mass o f fuel in the core (kg) W all thickness (m in) N um ber o f fuel assem blies (hexagouai) N um ber o f fuel rods in fuel assembly _ 10.88 4535 850 304 110 13.7 5130 80100 199.5 163 312 c Cooling system The reactor cooling system includes a prim ary side and a secondary one They are shown in Fig above M ain param eters are show n in Table below D.N Hai et aỉ / VNU Journal o f Science, M athem atics - Physics 21 (2 1) Ỉ Ỉ Ỉ - Ỉ 22 117 Table M ain param eters o f the cooling system _Prim ary loops (from to 4) _ H orizontal Steam Generator (Prim ary side) M ain C irculation Pum p (Capacity: mVh) 20000 Two m ain gate valves H ot leg length (m) 16.94 Cold leg length (m) 28.72 Pipe inner diam eter (m m ) 850 Pipe outer diam eter (mm) _ 990 Secondary loops (from Ĩ to 4) Feed w ater system Steam geneiator (secondary side) Steam lines, Turbine generator, C ondenser system d Steam generator V V ER reactors all use the horizontal steam generators Fig.5 shows a typical horizontal steam generator v«jd D nuurc N'ozzte B lw rknwi Noizie ' S«p«n&on lầtK MaaF««dn’aiKSu*vL’mi -GaiRkbov-iINonk i FeeT:ie Pressurized Light-W atcr R eactor Installations, Thermal Engineering, 54 (5) (2007) 343 [20] Lucia Bonavigo, M ario I)e Salve, Issues fo r Nuclear Pow er Plants Siearn Generators, Politecnico di Torino, Italy [2 ]N B T runov, B.I Lukascvich, D o Veselov, Yu G D ragunov, Steam generators-IIorizonial or Vertical (which type should be used in nuclear power plants with VVLR?), A tom ic Energy, 105 (3) (2008) 165 ... guyen T at T hang, A pplication o f some therm al hydraulic sim ulation softw ares to the m odeling, calculation and simulation of the thermal hydraulic system of the Da Lat nuclear lesearch reactor, ... Steam generator Feedw ater flowrate, kg/s Steam generator Steam generator Steam generator Steam generator Steam generator pressure, MPa Steam generator Steam generator Steam generator 5.86 Steam... various aspects related to nuclear reactors, nuclear pow er plants, nuclear safety etc A m ong those, the research and safety analysis based on numerical m odelings, calculations and sim ulations