DSpace at VNU: Thermal hydraulic system of a VVER-1000 nuclear reactor and numerical simulations

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 VV ER-1000 nuclear reactor

and numerical simulations

^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 T his 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 sim ulations

w ere conducted using the integrated VISA (V isual System A nalyzer) and R ELA P5 (R eactor

Excursion and Leak A nalysis Program , version 5) softw ares know n as V ISA _R ELA P5 Originally

R ELA P5 (a therm al hydraulic system code) and then recently VISA (a graphical user interface)

w ere d eveloped for the sim ulation o f the therm al hydraulic system o f W estern type pressurized

w ater reactors (P W R ) undergoing transients In V ietnam , research on the num erical sim ulation o f the therm al hydraulic system o f Da Lat nuclear research reactor and som e typical types o f PWRs

using R E L A P5 have long been carried out in our research group along w ith som e o f other research

institutions It should be noted that know ledge o f the V V ER 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 (N PP) in Russia Therefore this research has im portant 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, V ietnam 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: T herm al hydraulic, V ISA , RELA P5, N uclear R eactor, Pressurized W ater R eactor (P W R ), N uclear P ow er Plant (N PP), V V E R -1000, Ninh Thuan NPP

1 Introduction

A lthou gh n uclear pow er industry has experienced som e'serio us accidents in the past (C hernobyl accident in U kraine and Three M ile Island accident in the US) and recently the Fukushim a accident in

C o rresp o n d in g author: Tel.: (+84) 983384692

Em ail: ntthang@ im ech.ac.vn

11.1

112 D.N H ai et al / VNU Journal o f Science, M athem atics - P hysics 27 (2 0 1 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 o u r 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 2 r ‘ century.” [1],

Vietnam G overnm ent has approved plans to build 2 N PPs in V ietnam that are the and the 2"*'

N inh Thuan N PPs [2], It is therefore m andatory to develop hum an resources for nuclear industry (especially n uclear 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 m uch im portant as well

The best-estim ate therm al hydraulic sim ulation program s (e.g RELAP5 code [3]) have long been developed T hey have been step by step im proved to m odel m ore 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 lim ited 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 w ell on personal com puters As

a consequence, those sim ulations program are becom ing m ore and m ore popular H ow ever, 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 T herefore V ISA program was developed under the cooperation betw een KAERI (K orea A tom ic Energy Research Institute) and K H N P (K orea H ydro-N uclear Pow er), K orea to perform the tasks o f a G UI (G raphical U ser Interface) and to help users to exploit m ore effectively the therm al hydraulic sim ulation program s [4] VISA can be integrated with three therm al hydraulic simulation program s including M ARS, R ETRA N -3D and RELA P5 H ere w e use the integrated program VISA and RELAP5 w hich is called V ISA _R ELA P5 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 N PPs [6]

In the w orld now , the RELAP5 code (developed in the U S) is superior to other therm al hydraulic simulation codes in the nuclear industry and nuclear research T herefore in this research, we chose the V1SA_RELAP5 H ow ever 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 E nvironm ent Fluid D ynam ics, Institute o f M echanics, V ietnam A cadem y o f Science and Technology (V A ST) through the num erical

m odelings, calculations and sim ulations o f Da Lat nuclear research reactor in V ietnam and som e typical PW Rs using RELAP5 [7-14]

The application V ISA _RELA P5 to NPPs with Russian V V E R reactors is one o f the follow ing steps to contribute to raising our capability in thermal hydraulic research and safety analysis, and to

m astering the technology o f V V ER reactors that w ill 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 N inh 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 V ietnam and froữi abroad, our study w ould 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 som e extent, even in the w orld [16]), our research hopefully w ill 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 E R 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 w hen the reacto r operates at low pow er level o r in case o f em ergency etc.) [17, 18],

D.N H ai et a l / VNU Jo u rn a l o f Science M athem atics - Physics 27 (2 0 1 1) 111-122 113

2 Brief of VISA and RELAP5 codes

2.1 VISA

T he V ISA graphical user interface program was originally developed under the jo in t effort o f som e research institutions and energy industries in South K orea [4-6], T his program is designed to be integrated w ith som e therm al hydraulic sim ulation program s such as REL A P5, M ARS and R E T 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 program s 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) T he 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 in puư output 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 , th e 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 m im ics (G raphic and M im ic functions);

• V iew the output results through graphical w indow s, graphs o f variables (norm ally the variables o f tim e); m onitor the status o f trips (G raphic interface and T rip 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)

— ^ ^,/V ^0 ^

Fig 1 Graphical simulation o f a typical Western PWR (Mữnic iuactions) (figure from [5])

114 D.N H ai et a ỉ / V N V Jo u rn a l o f Science, M athem atics - Physics 27 (2 0 1 Ỉ) Ỉ I Ỉ - 122

Ị i t « « M r> »«i itm x

IIA m U m M Ì I'lMdttMM*!

VOID

* 140

Fig 2 G raphical presentation o f the PW R nodalization model and the calculated void in th e system during

calculation process (Pre-processor and Graphic functions) (figure from [5])

2.2 RELAP5

T heR E L A P S IS a best-estim ate transient sim ulation code for the sim ulation o f light w ater reactor coolant system during postTilated accidents Coupled behavior o f the reacto r co o lan t 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 therm al hydraulic system (i.e fuel rods, reactor core, control rods, reacto r vessel, pum p, heat conduction structures, pipe, valves, control system s etc.) O riginally the cod e was based on a hom ogeneous 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, hy drodynam ic model for transient sim ulation o f the tw o-phase system behavior The version used in this research is RELAP5/M OD2 w hich em ploys a full nonequilibrium , six-equation, tw o-fluid m odel

Stucy and applications o f R ELA P5/M O D 3.2 code was conducted in m an y o f our previous researches and w ill not be show n here D etails o f the system o f equations, the solution m ethods, the prograư flow chart, test calculations, and applications etc can be found in [3, 7-15, 17],

3 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 an y sim ilarities to

W estern PW Rs R ussia has m ade a lot o f effort to im prove m any aspects o f this reac to 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 w ith R ussian com panies to build a nuclear pow er p la n t in N in h T huan province (the r ‘ N P P o f V ietaam ) Therefore the reactor technology, n um erical m o deling , calculation, simulation and safety analysis o f the therm al hydraulic system o f V V E R n u clear pow er plant are urgently needed in V ietaam

3.Ỉ Structure o f the VVER reactor system

a O verview

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 3 below show s

D.N H ai et a! / VNU Journal o f Science M athem atics - Physics 27 (2011) I I I - I 2 2 115

the ty[:)ical therm al hydraulic system o f the V V ER-1000 reactor C om pared w ith W estern PW R (Fig 1) ones can see d e a rly som e differences V V ER reactors are equipped with isolation valves (Mam

G ate Valve - M G V ) and horizontal steam generator (Fig 2) Table 1 below show s som e o f the mam param eters o f the VVER-lOOOA^-338 reactor [18], In general, the V V E R nuclear reactor type has som e advantages o ver W estern PW Rs D etailed discussions can be referred further to in [19],

I'a b le 1 M ain param eters o f the VVER-lOOOA^-338 reactor

T herm al / Electric pow er 3000 M W th / 1000 M W e

C oolant P ressure (in the P rim 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)

15.7 M Pa

4

84800 mVh 289.7 “C

320.0°c

Fig 3 O verview o f the therm al hydraulic system o f a typical V V ER reactor (horizontal steam generator ) The m ain com ponents o f the therm al hydraulic system o f the V V E R reactor including (Fig 3):

1 - R eactor vessel

2 - R eactor core

3 - Control rods

4 - P re s s u r iz e r

5, 6 - H ot and cold legs (prim ary system )

7 , 8 - M a m G a te V a lv e s

9 - H orizontal S team G enerator

10 - M am circulation pum p (prim ary system)

11 - Steam line

12 - T u r b in e s

13 - C old leg (seco nd ary cooling system )

14 - C ontrol rod driving m echanism s

b R eactor vessel

The design o f the reactor vessel and the vessel’s m ain dim ensions are show n in Fig 4 M am param eters are show n in T able 2 below

116 D.N H ai et a ỉ / VNU Jo u rn a l o f Science, M athem atics - Physics 2 7 (2 0 Ĩ Ỉ) Ỉ Ỉ Ỉ - Ỉ 2 2

N t J / ‘ 1 I I ; !

r r m ầ _ l y Ị :

Fig 4 VVER-lOOOA^-338 reactor vessel

T able 2 M ain param eters o f the V V £R -1000A ^-338 reactor vessel

D iam eter, external on the cylindrical part o f the reactor (m in) 4535

N um ber o f fuel rods in 1 fuel assem bly _ 312

c C ooling system

T he reactor cooling system includes a prim ary side and a secondary one T hey are shown in Fig 3 above M ain param eters are show n in T able 3 below

D.N H ai et aỉ / VNU Journal o f Science, M athem atics - Physics 21 (2 0 1 1) Ỉ Ỉ Ỉ - Ỉ 22 117

Table 3 M ain param eters o f the cooling system

_Prim ary loops (from 1 to 4) _

H orizontal Steam G enerator (Prim ary side)

M ain C irculation Pum p (Capacity: mVh) 20000

Tw o m ain gate valves

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 E R reactors all use the horizontal steam generators F ig.5 show s a typical horizontal steam generator

1 v«jd

2 D nuurc N'ozzte

3 B lw rknwi Noizie ' S«p«n&on lầtK

6 MaaF««dn’aiKSu*vL’mi

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9 S tn ia N o s k

Ỉ 1 E f iif fie n n ' Fe« (» r» lB N o n fc

I: Acceũ Atriock Fig 5 H orizontal steam generator (figure from [20])

C om pared w ith vertical steam generators in W estern PW Rs, horizontal steam generators used in

V V E R reactors h ave som e typical advantages as show n below [20-21]:

• m oderate steam load and sim ple gravity-based steam separation m echanism ;

• m oderate velocity o f the cooling w ater in the second loop o f the steam generator thus preventing any dan ger o f vibrations o f the heat-exchange tube systeiTi;

• validated longtim e serviceability o f the steel tubes (the m axim um tim e o f operation is 38 years for

PG V -440 and 23 years for PG V -1000 type steam generators);

• vertical aư an g em en t o f the first-loop collectors, preventing accum ulation o f sludge deposits on their surfaces, thereby decreasing the danger o f corrosion dam age to the heat-exchange tubes in the region w here th e tubes are built into the tube sheet;

• larg er volum e o f w ater in the second loop, enabling the ability to cool the reactor via the steam generator in the case w here norm al and em ergency w ater feeding has stopped;

118 D.N H ai et aỉ / VNU Jo u rn a l o f Science, M aihem atics ' Physics 27 (2QỊ Ị) Ị 17 - / 2 2

in this steam generator design, It possible to m aintain an allow able con cen tratio n o f dissolved

im purities in the critical zones and increasing the reliability from the v iew p o in t o f corrosion effects;

horizontal aư angem ent o f the heat-exchange surface, enabling reliable natural circulation o f the first-loop coolant even w ith a m assive w ater level below the top row s o f the h eat-exchange tubes; convenient access to the tube sheet for servicing and checking from the first- and second-loop sides; there are no heat-exchange tubes at the bottom o f the housing, so that sludge is m ore easily rem oved through the purge system;

presence o f equipm ent for disconnecting the collectors from the m ain circulation pipelines, m aking

it possible to decrease the lime required to perform m aintenance w ork and lo increase the capacity utilization factor by perform ing w ork sim ullaneously on several steam generators and refueling the reactor

3.2 M odeling

Sim ilar to other num erical m odeling systems, the num erical sim ulation o f the therm al hydraulic system o f nuclear reactors using V ISA _RELA P5 requires the system be divided into control volum es,

o r thus nodalization m odel for every com ponent in the system (pum p, valve, pipe, reacto r core etc.)

B ased on the practical data o f the K alinin nuclear pow er plant [18], the n od alizatio n m odel for the num erical simulation has been developed (Fig 6)

As shown in Fig 6, nodalization model o f the VVER-lOOO/V-388 reactor shown in

V ISA _RELA P5 interface consists o f four circulation loops, these are a reactor vessel, four cooling loops, a pressurizer, four steam generators, steam lines, a Sleam collection and a steam storage system , safety valves, an exhaust valve o f steam generator system (Fig 6)

I'.- ■; I I ' l n <

; Ịíiị m -.1

Fig 6 N odalization m odel o i'th e them ial hydraulic system o f the VVER-lOOO/V-338 reactor

The M im ic functions o f V ISA _RELA P5 integrated system allow view ing the calcu lated therm al hydraulic param eters directly during the sim ulation process or after the sim ulation finishes (Replay

m ode) In this research, our M im ic model o f the V V E R -1000 reactor w as d evelo ped as show n

in Fig 7

D.N H ai el al / VNU Jo u rn a l o f Science M athem atics - Physics 27 (2 0 1 1 > 111-122 19

; t,', i r v .*.#• nr-t

f o Ù Li u-iL J *ti

' , j Fig 7 G raphical sim ulation o f the therm al hydraulic system o f the V V E R -1000 reactor using M im ic functions,

3.3 Results o f the num erical sim ulation o f the therm al hydraulic system o f the VVER-1000 reactor using VISA_RELAP5

C alculation and sim ulation w as initially carried out for the steady state o f the reactor operation at constant therm al p o w er 3000 M W th (the nom inal pow er o f the reactor) T he calculation shows that the system reaches steady slate after about 100 seconds Fig 8 show s reactor therm al pow er at steady state

3000 MWth

PL-?-::23

C V U lM (k

Fig 8 R eactor them ial pow er at steady state

Then m ajor therm al hydraulic param eters w hich w ere calculated using V ISA _R ELA P5 w ere com pared w ith the m easured data during the plant operation at the same operational condition (Table 4)

120 D.N H ai et al / VNU Jo u rn a l o f Science, M athem atics - Physics 27 (2 0 ! I) 111-122

Table 4 M ain therm al hydraulic param eters (m easured during the plant operation and calculated using

V IS A R E L A P 5 )

1 R eactor therm al pow er, M W th 2917.0 2917.5

2 C oolant flow rate, kg/s 18471.0 18471.2

3 Prim ary side pressure, M Pa 15.68 15.65

4 R eactor vessel inleưoutlet 326.0 328.6 pressure drop, KPa

5 C oolant tem perature at reactor vessel inleưoutlet, K

6 Feedw ater tem perature, K

7 Feedw ater flowrate, kg/s

8 Steam generator pressure, MPa

It is obvious that the calculated results using V ISA _RELA P5 w ith the nodalization model developed are in good agreem ent w ith the m easured data during the norm al operation o f the plant at nom inal therm al pow er and in steady state

The m odeling developed should be fully adequate for further studies on the tran sien t behaviors o f

V V E R -1000 reactors There exist a broad range o f therm al hydraulic safety issues related to the

V V E R reactors that need to be further investigated and addressed such as the R IA tran sient (Reactivity Lisertion A ccident), LO CA (Loss o f C oolant A ccident), LO FA (Loss o f Flow A ccident), FW LB (Feed

W ater Line Break accident), Loss o f O ffsite Pow er accident etc

4 Conclusions

The therm al hydraulic system o f a V V ER-1000 nuclear reactor has been investigated Basically the V V E R reactors are sim ilar to W estern type PW Rs Though they exhibit som e d ifferences and have