The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated
with RELAP5 code
Gaianê Sabundjian, Delvonei A. Andrade, Antonio Belchior Jr., Marcelo da Silva Rocha, Thadeu N. Conti,
Walmir M. Torres, Luiz A. Macedo, Pedro E. Umbehaun, Roberto N. Mesquita, Paulo H. F. Masotti, and Ana
Cecília de Souza Lima
Citation: AIP Conference Proceedings 1529, 151 (2013); doi: 10.1063/1.4804108
View online: http://dx.doi.org/10.1063/1.4804108
View Table of Contents: http://scitation.aip.org/content/aip/proceeding/aipcp/1529?ver=pdfcov
Published by the AIP Publishing
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The Behavior of ANGRA 2 Nuclear Power Plant
Core for a Small Break LOCA Simulated with
RELAP5 Code
Gaianê Sabundjian, Delvonei A. Andrade, Antonio Belchior Jr., Marcelo
da Silva Rocha, Thadeu N. Conti, Walmir M. Torres, Luiz A. Macedo,
Pedro E. Umbehaun, Roberto N. Mesquita, Paulo H. F. Masotti and Ana
Cecília de Souza Lima
Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN - SP)
Av. Professor Lineu Prestes, 2242
05508-000 São Paulo, SP - Brasil
Abstract. This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate
Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant
Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was
simulated with RELAP 5 code. The considered rupture area is 380 cm2, which represents 100%
of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity
of the reactor core.
Keywords: Small break LOCA, RELAP5, ANGRA 2.
PACS: 28.41-Fr
INTRODUTION
The objective of this work is to present the RELAP5/MOD3.2 gamma code [1]
behavior calculations of Angra 2 nuclear reactor core for a postulate loss of coolant
accident in the primary circuit, Small Break Loss of Coolant Accident (SBLOCA).
This accident and boundary conditions are described in detail in Chapter 15 of the
Final Safety Analysis Report of Angra 2 – FSAR [2]. The accident consists basically
of the total break of a pipe of the hot leg Emergency Core Cooling System (ECCS) of
Angra 2, which is a PWR reactor with four primary loops (10/20/30/40), FIGURE 1,
and power of 1,400MW(e). The rupture area is 380 cm2, which represents 100% of the
ECCS pipe flow area [3].
In this simulation, failure and repair criteria are adopted for the ECCS components,
in order to verify the system operation, in carrying out its function as expected by the
project to preserve the integrity of the reactor core and to guarantee its cooling, as
presented in the TABLE 1. SBLOCA accidents are characterized by a slow blowdown
in the primary circuit to values that the high pressure injection system is activated. The
thermal-hydraulic processes inherent to the accident phenomenon, such as hot leg of
ECCS vaporization and consequently core vaporization causing an inappropriate flow
distribution in the reactor core, can lead to a reduction in the core liquid level, until the
ECCS is capable to refill it.
XXXV Brazilian Workshop on Nuclear Physics
AIP Conf. Proc. 1529, 151-154 (2013); doi: 10.1063/1.4804108
© 2013 AIP Publishing LLC 978-0-7354-1154-8/$30.00
151
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BREAK
ECCS
(HOT
LEG)
10
REACTOR VESSEL
086
6
ECCS IN HOT LEGS OF
THE PRIMARY LOOPS
(10//30/40)
088
005
5
4
090
3
284
2
1
2
2
to hot leg
3
3
MJ - 079
100
4
1
from cold leg
1
eccs10
SJ - 077
4
160
066
4
460
076
MJ - 075
010
4
3
6
2
2
7
2
1
1
8
3
300
400
5
1
SJ - 097
013
MJ - 053
BYP
11
11
10
10
9
9
9
8
8
8
8
036
10
BTC4
11
050
10
BTC3
11
12
048
BTC2
10
12
046
11
12
6
6
6
6
10
7
9
9
9
9
9
8
8
8
8
8
9
044
11
12
042
VECQ
11
12
12
040
11
SJ - 037
12
038
5
12
CM
4
ECQ
3
BTC1
360
10
6
6
6
6
6
11
5
5
5
5
5
5
5
5
12
4
4
4
4
4
4
4
4
13
3
3
3
3
3
3
3
3
14
2
2
2
2
2
2
2
2
15
1
1
1
1
1
1
1
1
10
10
MJ - 051
260
200
eccs4
0
eccs30
eccs20
095
SJ - 011
16
020
17
017
015
MJ - 051
038 - 040
038 - 048
044 - 046
038 - 044
038 - 050
048 - 050
038 - 046
040 - 042
REACTOR
CORE
FIGURE 1. ANGRA 2 vessel RELAP5 code nodalization.
152
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TABLE 1. Injection by the ECCs for rupture of the Loop 20 hot-leg injection line
ECCs
Components
Safety Injection
Pumps
Accumulators
Residual
Heat
Removal Pumps
Break:
SF:
RC:
Injection
Loop 10
hot
1
cold
_
1
Loop 20
hot
Break
1
cold
_
Loop 30
hot
SF
cold
_
Loop 40
hot
RC
cold
_
1
1
1
1
Break
1
Break
1
SF
RC
Injected coolant lost via the break
Single failure of diesel engine
Diesel engine down for repairs
RESULTS AND CONCLUSIONS
Results obtained with RELAP5 to the core of ANGRA2, for the considered
SBLOCA, are presented in the FIGURES 2, 3, 4 and 5:
tempf 42050000
200
350
300
Temperature ( °C)
Pressure (bar)
160
120
80
40
RELAP5
250
200
150
100
50
0
FSAR
0
-40
0
0
200
400
600
800
1000
200
400
1200
600
800
1000
1200
Time (s)
Time (s)
FIGURE 3. Temperature in the reactor core
RELAP5
FIGURE 2. Pressure in the reactor core
(RELAP5 and FSAR)
rktpow 0
800
4000000000
700
Total Reactor Power (W)
Temperature ( °C)
600
500
400
300
RELAP5
200
100
3000000000
2000000000
1000000000
FSAR
0
0
0
200
400
600
Time (s)
800
1000
0
1200
200
400
600
800
1000
1200
Time (s)
FIGURE 4. Maximum Temperature in the
hot fuel rod (RELAP5 and FSAR)
FIGURE 5. Total reactor power - RELAP5
FIGURES 2 and 3 show the pressure and the temperature in the core cooling
channels obtained with RELAP5 code. FIGURE 2 also compares the pressures
153
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obtained from the simulation with RELAP5 and FSAR that showed to be in a
reasonable agreement.
FIGURE 4 shows that the core cooling by ECCS was enough to keep the fuel
cladding below its melting temperature (1200 0C). Although the results were not as
expected when compared to the FSAR, the ECCS efficiency was verified for this
accident.
The reactor control system simulated in this work was able to shutdown the reactor
in the safety set point of ANGRA 2, FIGURE 5.
Results presented in this paper showed the correct actuation of the ECCS
guaranteeing the integrity of the reactor core.
Further work will be developed to better understand the FSAR differences and
guide us to improve the RELAP5 model.
REFERENCES
1.
2.
3.
RELAP5/MOD3.3 Code Manual, NUREG/CR-5535/Rev1, IDAHO LAB. SCIENTECH Inc.,
Idaho (2001).
ETN, Final Safety Analysis Report – Central Nuclear Almirante Álvaro Alberto – Unit 2,
ELETRONUCLEAR S/A., Doc. Ident. MA/2-0809.2/060000, Rev. 3, Abril 2000.
Andrade, D.A. & Sabundjian, G., “Qualificação a nível transiente da nodalização a2nb03c:
Acidente de SBLOCA de 380 cm2 da linha do sistema de resfriamento de emergência do núcleo
(SREN), conectada à perna quente”. Instituto de Pesquisas Energéticas e Nucleares –
IPEN,Relatório Técnico., julho 2001.
154
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