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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
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CONTENTS
• Introduction of Safety Analyses Department• VVER-440 description• Development of 3D model for VVER-440/213• Model validation • Conclusion
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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• 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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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• 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)
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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Fig . 2: Main steam system nodalization
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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Fig . 3: Feed water system nodalization
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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Fig. 7: Core model
349 fuel assemblies
31 TH channels
6 reflector channels
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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Fig. 10 : Temperatures in core channels 6 and corresponding fuel assemblies
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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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Fig. 11 : Temperatures in core channels 13 and corresponding fuel assemblies
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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
Nuclear Research Institute Řež plc 2010 RELAP5 International User’s Seminar
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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
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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 .