DEVELOPMENT OF RELAP5-3D MODEL FOR VVER-440 REACTOR

1

Nuclear Research Institute Řež plc

DEVELOPMENT OF RELAP5-3D MODEL FOR

VVER-440 REACTOR

2010 RELAP5 International User’s Seminar

West Yellowstone, MontanaSeptember 20-23, 2010

Marek Benčík, Jan Hádek

Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar

2

CONTENTS

• Introduction of Safety Analyses Department• VVER-440 description• Development of 3D model for VVER-440/213• Model validation • Conclusion


3

1. NRI INTRODUCTION

• Main activities:– Thermal hydraulic and neutron kinetics

calculations– Development of advanced analytical methods– Expert missions

• NPP Type– VVER 440/213 Dukovany– VVER 1000 Temelín– PWR


4

• Thermal hydraulic and neutron kinetics calculations for:– Safety Analysis Report (SAR)– Pressurized Thermal Shock (PTS) analyses– Equipment qualification– Computer codes validation (e.g. under umbrella of

OECD)– Verification of Emergency Operation Procedures– Accidents at low power and shutdown– Probabilistic Safety Assessment


5

• Development of advanced analytical methods– Best estimate analyses with considered uncertainty

of BE computer codes models, correlations and input data

– Prediction of CHF use of CFD codes. New model of CHF calculation base on microstructure of process.

– Two phase CFD computer codes. • Expert missions include:

– Support to the Czech nuclear power plants– Support to the regulatory bodies (Czech, Ukraine)– Consideration of new nuclear facilities– Participation in the development of advanced

nuclear power plants (for example Generation IV)


6

2. VVER-440 (NPP DUKOVANY)

VVER-440 description:• Thermal power: 1444 MW• Primary pressure: 12,36 MPa• Number of fuel assemblies: 349• Loops: 6 • Horizontal steam generators


7

The RELAP5/MOD3 input model for VVER-440 created in NRI Řež (P.Král, L. Krhounková) was used as a base for the RELAP5-3D input deck.

Main characteristics of input model are following: The reactor vessel is described by two 3D multid object and bundle of 1D channels (core). All 6 loops of reactor coolant system are fully modeled, as well as the pressurizer system, ECCS system, main steam system (MSS) and feed water (FW) system lines, all relevant heat structures, control and protection systems.

3. DEVELOPMENT OF MODEL


8

Fig. 1 : Nodalization of reactor coolant system

Fig. 1 : Nodalization of reactor coolant system

2

9

9

1

854

3

6 7

2

3

1

5

8

9

1011

13

14

15

16

5

6

7

8

6

12

4

7

10


9

Fig . 2: Main steam system nodalization


10

Fig . 3: Feed water system nodalization


11

Fig. 4: Reactor pressure vessel nodalization Fig. 5: Fuel assembly

125

288258228

UP 088

CL

041CORE

438418HA

188158

077

HL

408428

HA

DC + LP035

100130160200230260

7

8

9

3

14

15

16

2

345

6

7

4

89

10

1

1

2

6


12

Neutronic Library:

KASSETA HELIOS

Program DYLIE

RELAP5-3D

Library: CSLIBR

RELAP5-3D

Core model

Subroutine userxs

calling subroutines:

param

rods

record

CSLIBR

D, a, f, f, s In kinetics node

Tm, m, Tf, cb

idrcrd, crdfr, valusr,

time, dt, mode

Fig. 6: Neutronic data preparation


13

Fig. 7: Core model

349 fuel assemblies

31 TH channels

6 reflector channels


14

4. ANALYSED SCENARIOS

RELAP5-3D with 3D TH and NK is particularly suitable for analyses of cases with non uniform power generation in core. For the time being the following cases were calculated for VVER 440:

• Steam line break (full power, HZP)• Malfunction of feed water system (HZP)• Boron dilution (HZP)• AER 6 benchmark

Correct prediction of mixing in the pressure vessel is essential


15

0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

Fig. 8: Malfunction of feed water system, HZP 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8

Loop 1

Loop 6Loop 5

Loop 4

Loop 3 Loop 2

240,0

241,0

242,0

243,0

244,0

245,0

246,0

247,0

248,0

249,0

250,0

251,0

252,0

253,0

254,0

255,0

0,0 100,0 200,0 300,0 400,0 500,0Time [s]

Tem

per

atu

re [

°C]

T 1 T 2 T 3 T 4 T 5 T 6 T 7 T 8 T 9 T 10 T 11 T 12 T 13

T 14 T 15 T 16 T 17 T 18 T 19 T 20 T 21 T 22 T 23 T 24 T 25 T 26

T 27 T 28 T 29 T 30 T 31


16

5. MODEL VALIDATION

Mixing of coolant in reactor during asymmetrical cool down

Initial conditions:• Reactor operated on HZP• 6 MCP in operation• Average primary temperature 260°C• Secondary pressure 4.56 MPa


17

Scenario description:• The selected cold leg is cooled down (~5°C)

by steam reduction station• Temperatures in core, cold and hot legs are

measured• Test is repeated for all the loops

RELAP5-3D model:• Only pressure vessel is modeled


18

252,0

253,0

254,0

255,0

256,0

257,0

258,0

259,0

260,0

261,0

262,0

38000,0 40000,0 42000,0 44000,0 46000,0 48000,0

Time [s]

Te

mp

era

ture

[°C

]

Relap (channel č. 2) measurement 42 measurement 82

Fig. 9 : Temperatures in core channels 2 and corresponding fuel assemblies


19


252,0

253,0

254,0

255,0

256,0

257,0

258,0

259,0

260,0

261,0

262,0

38000,0 40000,0 42000,0 44000,0 46000,0 48000,0

Time [s]

Te

mp

era

ture

[°C

]

Relap (channel č. 6) measurement 96 measurement 97 measurement 79 measurement 80

measurement 67 measurement 68 measurement 78 measurement 64 measurement 65


20


252,0

253,0

254,0

255,0

256,0

257,0

258,0

259,0

260,0

261,0

262,0

38000,0 40000,0 42000,0 44000,0 46000,0 48000,0

Time [s]

Te

mp

era

ture

[°C

]

Relap(channel č.13) measurement 150 measurement 186 measurement 187 measurement 208

measurement 5 measurement 6 measurement 17


21

Fig. 12 : Temperatures in hot legs

255

256

257

258

259

260

261

38000 40000 42000 44000 46000 48000 50000

Time [s]

Te

mp

era

ture

[°C

]

TCH( 1) TCH( 2) TCH( 3) TCH( 4) TCH( 5) TCH( 6) Relap 1 Relap 2

Relap 3 Relap 4 Relap 5 Relap 6


22

6. CONCLUSIONS AND FUTURE WORK

• 3D TH and NK model of VVER 440/213 has been developed and improved during last decade in NRI Řež

• The paper presents our last validation effort focused on mixing in reactor vessel.

• The results of 3D calculation are in good agreement with measured data

• The original complex model of VVER 440 is too complicated, unsuitable for sensitivity and uncertainty studies

• Optimalization and simplification are needed


23

REFERENCES

• Král P.: Assessment of RELAP5 and Verification of Modelling Methods for VVER-Type Reactor Analysis. Paper for the RELAP5 International Users Seminar. Boston. July 1993.

• Macek J., Muhlbauer P., Krhounková J., Král P., Malačka M.: Thermal Hydraulic Analyses of NPPs with VVER-440/213 for the PTS Condition Evaluation. NURETH-8. 1997.

• Hádek J., Král P.: Final Results of the Sixth Three-Dimensional AER Dynamic Benchmark Problem Calculation. Solution of Problem with DYN3D and RELAP5-3D Codes. 13th Symposium of AER on VVER Reactor Physics and Reactor Safety, Dresden, September 2003 .

DEVELOPMENT OF RELAP5-3D MODEL FOR VVER-440 REACTOR

Documents

Transcript of DEVELOPMENT OF RELAP5-3D MODEL FOR VVER-440 REACTOR