iea-r1 primary and secondary coolant piping systems ... · development of the primary and secondary...
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2013 International Nuclear Atlantic Conference - INAC 2013
Recife, PE, Brazil, November 24-29, 2013 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN
ISBN: 978-85-99141-05-2
IEA-R1 PRIMARY AND SECONDARY COOLANT PIPING SYSTEMS
COUPLED STRESS ANALYSIS
Gerson Fainer1, Altair A. Faloppa
1, Carlos A. Oliveira
1 and Miguel Mattar Neto
1
1 Instituto de Pesquisas Energéticas e Nucleares - IPEN / CNEN - SP
Av. Professor Lineu Prestes, 2242
05508-000, São Paulo, SP, Brazil
[email protected]; [email protected]; [email protected]; [email protected]
ABSTRACT
The aim of this work is to perform the stress analysis of a coupled primary and secondary piping system of the
IEA-R1 based on tridimensional model, taking into account the as built conditions. The nuclear research reactor
IEA-R1 is a pool type reactor projected by Babcox-Willcox, which is operated by IPEN since 1957. The
operation to 5 MW power limit was only possible after the conduction of life management and modernization
programs in the last two decades. In these programs the components of the coolant systems, which are
responsible for the water circulation into the reactor core to remove the heat generated inside it, were almost
totally refurbished. The changes in the primary and secondary systems, mainly the replacement of pump and
heat-exchanger, implied in piping layout modifications, and, therefore, the stress condition of the piping systems
had to be reanalyzed. In this paper the structural stress assessment of the coupled primary and secondary piping
systems is presented and the final results are discussed.
1. INTRODUCTION
IEA-R1 reactor is a pool type, light water moderated and graphite reflected research reactor.
Its first operation was in 1957 in the ancient Institute of Atomic Energy, currently named
IPEN (Nuclear and Energy Research Institute), in Sao Paulo. The original design was
developed by the American company Babcock & Wilcox, and the reactor started with 2 MW
power [1].
Since its start-up in 1957, the IEA-R1 has been modified for several reasons such as
replacement of aged structures and components, upgrading of systems, and adequacy to
safety codes and standards that have changed during the years. The main modifications up to
2012 are listed bellow: [2], [3]
1971: - Changes in reactor building ventilation system.
1974: - Duplication of the primary coolant system to redundancy;
- Introduction of flywheels on the primary centrifugal pumps;
- Installation of a 16
N decay tank to reduce operational dose.
1976: - Replacement of original instrumentation and control operation console;
- Replacement of the original control drive mechanism;
- Changes in the pneumatic system.
INAC 2013, Recife, PE, Brazil.
1978: - Replacement of the original ceramic liner of the pool walls by steel liner;
- Installation of an auxiliary bypass system for emergency water supply.
1987: - Separation of the reactor building area into radiological hot and cold rooms;
- Introduction of access chambers;
- Introduction of vertical duct for irradiated material transfer;
- Installation of a new cooling tower;
- Main maintenance of the emergency generators.
1991: - Construction of a radiation shielding for the resins columns in the pool water
purification and treatment system.
1995: - Power upgrade from 2 MW to 5 MW.
2001: - Main maintenance of the cooling towers.
2002: - Main maintenance of primary pumps;
- Installation of beryllium reflectors.
2005: - Replacement of a heat exchanger (TC1A).
2007 - Installation of a new pneumatic system
2012: - Withdrawal of check valves of the secondary system;
- Replacement of secondary pumps.
The modifications indicated by these processes: “duplication of primary coolant system to
redundancy”, “installation of decay tank” and “replacement of a heat exchanger” in the
Primary Circuit, and “installation of a new cooling tower”, “withdrawal of check valves” and
“Replacement of secondary pumps” in the Secondary Circuit caused changes in piping
layout.
This paper is the outcome of the structural analysis of the IEA-R1 coupled Primary and
Secondary Circuits piping after these modifications. In order to develop this work it was
carried out the following schedule:
assessment of piping documents;
verification of the as-built condition of the primary and secondary piping systems;
construction of a 3D solid model of the primary and secondary piping systems;
preparation of the isometric drawings for stress calculation;
development of the primary and secondary piping structural model, processing and post-
processing of results, and analysis of the obtained results.
The piping analysis was performed by using the program CAESAR II [4], a very well know
piping system analysis software, which has become the standard for piping stress analysis.
The stress analysis was performed as follow: the Primary Circuit Piping was classified as
nuclear safety class 2, according to the ASME code, section III, subsection NC [5] and the
Secondary Circuit Piping was classified as non nuclear, according to the ASME code B31.1
[6].
2. PRIMARY AND SECONDARY CIRCUIT DESCRIPTION
The figure 1 shows a simplified flowchart of the Primary and Secondary Systems of the IEA-
R1 reactor.
INAC 2013, Recife, PE, Brazil.
The primary circuit provides the water to cool the reactor core. It is an open circuit, since the
refrigeration is provided by the own pool water in the reactor. In this system, the water passes
through the plates of the reactor core and the header located at the bottom of the pool, coming
out of the pool through the piping nozzle. Then, passes through a decay tank of 16
N and it is
pumped by pumps B1A and/or B1B, to the respective heat exchanger TC1-A and/or TC1-B,
where it's cooled, returning, then, to the pool through a diffuser plugged in the piping nozzle.
The system has redundant pumps, heat exchangers and valves.
The secondary circuit provides the water to cool the heat exchanger TC1-A and/or TC1-B. It
is also an open circuit, since the refrigeration is provided by cooling towers TR1-A and/or
TR1-B. The returning water is pumped by B102-A and/or B102-B to the respective heat
exchanger TC1-A and/or TC1-B.
Figure 1. Simplified flow chart of the IEA-R1 Primary & Secondary Circuit
To better analyze the system, in a comprehensive manner, a three-dimensional solid model of
the complete primary and secondary circuit was built using the program "SOLIDWORKS"
[7]. This work was initiated with the preparation of a complete isometric drawing of all
components of the primary and secondary circuits based on the as-built condition. Then, with
the objective to show the whole system, to recheck the dimensions to guarantee that the as-
built is correct, and to allow the preparation of the isometric for stress calculation, a 3D solid
model was done. The resultant 3D solid model is shown in the figure 2.
Decay
Tank CP-VGV-04
Pump B1-A
CP-VGV-03
CP-VGV-02
Pump B1-B
diffuser
core POOL
CP-VRT-01
CP-VRT-02
CP-VGV-06
CP-VRT-05
CS-VGV-04
Pump B102-B
CS-VGV-07
CS-VGV-02
Pump B102-A
CS-VGV-06
CP-VIS-02
CP-VIS-01
Heat
Exch.
TC1A
CP-VGV-01
CP-VGV-08
CP-VGV-07
Heat
Exch.
TC1B CP-VIS-03
CP-VIS-04
CP-VGV-09
CS
-VG
V-1
4
CS
-VG
V-1
1
CS
-VG
V-1
5
CS-VGV-10
CS-VGV-03
Cooling
Tower
TR-1B
Cooling
Tower
TR-1A
Primary
Circuit
Secondary
Circuit
CS-VGV-01
CS-VGV-05
INAC 2013, Recife, PE, Brazil.
Figure 2. 3D Model of the IEA-R1 Primary & Secondary Circuit
Cooling Tower
Decay Tank
Heat Exchanger
Primary Pump
Secondary Pump
Pool
Core
Suspension Frame
- Primary Circuit Piping
- Secondary Circuit Piping
INAC 2013, Recife, PE, Brazil.
3. PIPING MODEL DESCRIPTION
For piping analysis purpose, the primary and secondary circuits of the reactor IEA-R1, as
described in item 2, were modeled as a whole, encompassing piping and equipment coupled
together. The program CAESAR II [4] was used for the analysis. Calculation model was
developed and nodes and elements are shown in the isometric for stress calculation presented
in figures 3 and 4.
The following theoretical premises were adopted:
The Primary Circuit Piping was considered as Nuclear Safety Class 2, according to the
ASME code, section III, subsection NC [5], and the Secondary Circuit Piping as non
nuclear, according to ASME B31.1 [6];
To build the model, straight pipe and bend elements were used, with three DOF (Degrees
Of Freedom) for translation and three DOF for rotation;
The valves were modeled as rigid elements;
The components were modeled as equivalent beams;
The piping supports were modeled by adopting a restraint stiffness of 1 x 105 N/mm to
the restrained direction;
The inlet and outlet pool nozzles, the heat exchangers nozzles, the decay tank nozzles, the
cooling towel nozzles, and the suction and discharge pump nozzles were modeled by
adopting a restraint stiffness of the 1 x 107 N/mm for the translational DOF’s and 1 x 10
12
Nmm/rad for rotational DOF’s;
Components stress-intensification factor’s (SIF), were modeled as follows: “tees” were
modeled as “welding tee” and “reductions” were modeled according to their geometry.
The Calculation Model will be valid whether the following premises were true:
The piping, valves, tees, flanges and reductions must be without defects, it means: there is
no denting, no pitting, no misalignments, no out of roundness and nor any corroded points
on external or internal surface of the piping or fittings;
The attachments, welded or threaded, that connects piping to piping, piping to equipment
and piping to fittings (valves, tees, flanges, reductions) must be without defects, it means:
there is no denting, no pitting, no misalignments, no out of roundness and nor any
corroded points on external or internal surface of the fittings;
INAC 2013, Recife, PE, Brazil.
1780
980
2110
3030
6501
300
386
1690
15
46
1760
31°V
360
41°V
400
1838
93
0
2710
985
54
0
1800
2210
3840
1390
2360
340
225
Po
ol In
let
DN
10"
880
DN
10"
NO
TES
:
1)D
IM
EN
SIO
NS
IN
M
M2
)ELEV
ATIO
N I
N M
ETER
DN
10"
900
45°V
2895
45°V
8714
914
EL.
75
7,3
60
1755
1275
2333
1414
45°V
737
13076
45°H
1619
1178
Po
ol O
utl
et
DN
16"
DN 12"
Pu
mp
B1
-B
Pu
mp
B1
-A
1138
668
DN
16"
305
220
305
130
160
220
54
0
45°V
183
Decay T
ank
Heat
Exchanger
TC1-B
-CBC
Heat
Exchanger
TC1-A
-IESA
V-0
1
V-0
3
V-1
1
V-1
2
V-1
3
V-0
7
V-0
4
V-0
9
V-1
5V
-1
4
V-1
0
V-0
6
V-0
5
V-0
8
13
50
13
40
13
30
13
20
12
90
12
60
12
50
12
20
78
0
77
0
81
0
72
0
71
0
68
0
67
06
40
11
50
11
20
10
00
99
09
80
94
5
93
0
90
08
90
82
0
84
0
86
0
11
00
59
0
58
0 57
0
55
0
50
5
38
0
43
0
10
50
10
40
35
0
34
0
30
0
29
0
28
02
50
24
0
17
0
14
0
13
0
11
090
60
50
20
10
V-0
2
Valv
e L
ist
Nu
mb
er
TA
G
V-0
1
V-0
2
V-0
3
V-0
4
V-0
5
V-0
6
V-0
7
V-0
8
V-0
9
V-1
0
V-1
1
V-1
2
V-1
3V
-1
4
V-1
5
CP
-V
IS
-0
2
CP
-V
IS
-0
1
CP
-V
GV
-0
1
CP
-V
GV
-0
3
CP
-V
RT-0
1
CP
-V
GV
-0
5
CP
-V
GV
-0
2
CP
-V
RT-0
2
CP
-V
GV
-0
6
CP
-V
GV
-0
4
CP
-V
IS
-0
3
CP
-V
IS
-0
4
CP
-V
GV
-0
9
CP
-V
GV
-0
8
CP
-V
GV
-0
7
EL.
75
6,3
60
EL.
75
8.4
95
EL.
75
8,5
30
EL.
75
6,3
60
EL.
75
8,1
00
EL.
75
9,0
99
EL.
75
8,8
60
EL.
75
8,2
68
EL.
75
9,3
85
EL.
75
8,3
55
EL.
75
9,5
54
45°V
EL.
75
8,3
53
EL.
75
8,8
05
EL.
75
8,0
43
EL.
75
9,5
54
Figure 3. Isometric drawing for stress calculation of the IEA-R1 Primary Circuit
INAC 2013, Recife, PE, Brazil.
NO
TE
S:
1)D
IME
NS
ION
S I
N M
M
2)E
LE
VA
TIO
N I
N M
ET
ER
Va
lve L
ist
Num
ber
TA
G
V-0
1
V-0
2
V-0
3
V-0
4
V-0
5
V-0
6
V-0
7
V-0
8
V-0
9
V-1
0
V-1
3
V-1
4
V-1
5
CS
-VG
V-0
1
CS
-VG
V-0
2
CS
-VG
V-0
3
CS
-VG
V-0
4
CS
-VG
V-0
5
CS
-VG
V-0
6
CS
-VG
V-0
7
CS
-VG
V-0
8
CS
-VG
V-0
9
CS
-VG
V-1
0
CS
-VG
V-1
3
CS
-VG
V-1
4
CS
-VG
V-1
5
?
?
?
?
?
?
Bocais
de
Entr
ada/S
aíd
a
Bom
ba B
102
-A (
DN
10")
EL.
758,3
7
Bo
cais
de
Entr
ada/S
aíd
aB
om
ba B
102
-B (
DN
10")
EL.
75
8,3
7
TR
1-A
TR
1-B
Alp
ina
BW
1221
850
4880
6610
6462
11303
900
1027
900
1550
1000
380
1540
850
(DN
10")
3060
2103
670
400
5350
300
2542
3080
500
1624
9665
600
2020
1260
770
1136
523
45°V
EL
. 7
61,7
3
EL. 7
61
,73
2543
4060
14550
11000
7252 tip
.
1500
2350
14000
6543
3004 tip
.
22
°30'H
11800
980
3728
10150
5950
4082
538
TC
2-C
BC
TC
1- IE
SA
12825
2140
V-0
4
DN
10"
EL
. 7
58
,834
EL
. 7
58
,428
538
1316
V-0
5
V-0
2V
-03
V-0
6
V-0
7
V-0
1
V-1
4
V-1
5
V-1
1 V-1
0
V-1
2
V-1
3
2900
2920
2750
2720
2180
2200
2220
2960
2570
3050
2980
2260
2680
3650
3630
3550
2590
2620
2550
2950
2580
3570
2250
EL
. 7
58
,370
EL
. 7
58
,370
DN
18"
DN
16"
DN
12"
DN
10"
DN
12"
DN
12"
DN
10"
DN
10"
DN
8"
DN
8"
2280
2310
30702
340
2380
3160
2440
2470
3200
3370
3440
3490
3530
3220
3280
3110
2540
3410
2010
2020
2050
2035
Figure 4. Isometric drawing for stress calculation of the IEA-R1 Secondary Circuit
INAC 2013, Recife, PE, Brazil.
3.1. General Data
The materials used in the analysis are: a stainless steel for the Primary Circuit Piping and
carbon steel for the Secondary Circuit Piping. The physical and mechanical properties of the
piping materials are shown in table 1 and the geometric properties of the piping in table 2.
Table 1: Materials Properties ([6], [8])
Material Thermal Expansion
(mm/mmºC)
Modulus of Elasticity
(N/mm2)
Stress (N/mm2)
Allowable Yield
SA-358 TP304
(stainless steel) 15.6x10
-6 195100 137.0 206.0
A 671 C65
(carbon steel) 11.8x10
-6 201000 112.5 241.5
Table 2: Piping Cross Sections Properties (mm)
Primary Circuit Secondary Circuit
DN Outlet diameter Thickness DN Outlet diameter Thickness
250 274.64 3.175 200 219.07 8.18
300 311.15 3.175 250 273.05 9.27
350 355.60 3.960 300 323.85 9.52
400 410.37 3.581 400 406.04 9.52
450 457.20 9.52
The operation modes of the IEA-R1 primary and secondary circuits are shown in table 3.
Table 4 lists the process data for pressure and temperature applied in the analysis.
Table 3: Operation Modes
Primary Circuit Secondary Circuit
Operation Pump Heat Exchanger Pump Cooling Towel
Mode B1A B1B TC1A TC1B B102A B102B TR1A TR1B
#1 On Off On Off On Off On Off
#2 Off On Off On Off On Off On
#3 On On On On On On On On
Table 4: Process condition [2]
Primary Circuit Secondary Circuit
Pressure (N/mm2) 0.69 0.518
Temperature (ºC) 43.9 37.3
INAC 2013, Recife, PE, Brazil.
3.2. Load cases
In order to perform the piping stress analysis, the following load cases were defined:
DW - (Dead Weight) - Pipe, fluid, flanges and valves weight,
PD - (Design Pressure) - Maximum pressure applied to the lines (see table 4),
Th - (Thermal Expansion) - Three Thermal load cases (see table 3):
Th1 Operation Mode #1;
Th2 Operation Mode #2;
Th3 Operation Mode #3.
3.3. Stress analysis
Piping stress analysis was performed according to ASME code. As stated before, the primary
circuit piping was classified as nuclear safety class 2 and analyzed by section III subsection
NC [5], and the secondary circuit piping was classified as non-nuclear and analyzed by B31.1
[6].
The ASME code provides the criteria for the stress evaluation of components and piping
based on two categories of loading. Such categories are related to non-self-limiting and self-
limiting stresses.
A non-self-limiting stress is produced by the mechanical load that causes primary stresses,
which are responsible for the equilibrium of the system. When they exceed the material’s
yield strength it could result in failure or gross distortion. Pressure and weight are typical
mechanical loads.
A self-limiting stress is generated by restraining the deformation of the system and, generally,
produces secondary stresses that can lead to ratcheting. Thermal expansion is a typical
example of this load.
Table 5 shows ASME code equations according to plant condition, failure mode and category
of the loads.
Table 5: ASME code Equations
Condition Failure Mode
Equation
Stress categorization NC B31.1
Design plastic collapse 8 11B Primary
Service ratcheting 10 13A Secondary
To prevent gross rupture of the system, the general and local membrane stress criteria must
be satisfied. This is accomplished by meeting equation (8) of NC-3652 of the ASME code [5]
INAC 2013, Recife, PE, Brazil.
and equation (11B) of 104.8.1 of the ASME B31.1 [6]. These equations require that primary
stress intensity resulting from design pressure; weight and other sustained loads meet the
stress limit for the design condition.
To prevent ratcheting, the secondary stress criteria must be satisfied. This is accomplished by
meeting equation (10) of NC-3653.2 of the ASME code [5] and equation (13A) of 104.8.3 of
the ASME B31.1 [6]. These equations require that secondary stress intensity resulting from
thermal expansion meet a level A and B service limits.
The equations for design and service conditions for ASME – NC and ASME – B31.1 are
summarized in tables 6 and 7.
Table 6: Equations of ASME Code Section III subsection NC [5]
CONDITION: Design CONDITION: Thermal Expansion
(Equation 8) ha
n
oS
Z
MB
t
PDB 5.1
221 (Equation 10) A
c SZ
Mi
where:
B1, B2 → primary stress indices;
oD → outside diameter of pipe;
tn → nominal wall thickness;
P → internal design pressure;
Ma → resultant moment loading on cross
section due to weight and other sustained
loads;
Z → section modulus of pipe;
Sh → basic material allowable stress at design
temperature.
where:
i → stress intensification factor;
cM → range of resultant moments due to
thermal expansion;
SA → tensão admissível p/ cargas cíclicas:
hcA SSfS 25.025.1 ;
Sc → basic material allowable stress at room
temperature (20º C);
f → stress range reduction factor for cyclic
loads
f = 1 (N ≤ 7000 cycles).
Table 7: Equations of ASME Code - B31.1 [6]
CONDITION: Design CONDITION: Thermal Expansion
(Equation 11B) ha
n
oS
Z
Mxi
t
PD 0.1)75.0
4( (Equation 13A) A
c SZ
Mi
where:
i → primary stress intensification factor;
oD → outside diameter of pipe;
tn → nominal wall thickness;
P → internal design pressure;
Ma → resultant moment loading on cross section
due to weight and other sustained loads;
Z → section modulus of pipe;
Sh → basic material allowable stress at design
temperature.
where:
i → stress intensification factor;
cM → range of resultant moments due to
thermal expansion;
SA → allowable stress range for expansion
stresses:
hcA SSfS 25.025.1 ;
Sc → basic material allowable stress at room
temperature (20º C)
f → stress range reduction factor for cyclic
loads
f = 1 (N ≤ 7000 cycles).
diffuser
Pump B1-B
Decay Tank
Pump B1-A
CP-VGV-02
CP-VGV-03
Heat Exchanger
TC1B
Heat Exchanger
TC1A
CP-VGV-04 CP-VRT-01
CP-VRT-02
CP-VGV-06
CP-VGV-05
CP-VIS-02
CP-VIS-01
CP-VGV-01
CP-VGV-08
CP-VGV-07
CP-VIS-03
CP-VIS-04
CP-VGV-09
POOL core
diffuser
Pump B1-B
Decay Tank
Pump B1-A
CP-VGV-02
CP-VGV-03
Heat Exchanger
TC1B
Heat Exchanger
TC1A
CP-VGV-04 CP-VRT-01
CP-VRT-02
CP-VGV-06
CP-VGV-05
CP-VIS-02
CP-VIS-01
CP-VGV-01
CP-VGV-08
CP-VGV-07
CP-VIS-03
CP-VIS-04
CP-VGV-09
POOL core
INAC 2013, Recife, PE, Brazil.
4. STRESS ANALYSIS RESULTS
The piping calculation model, built in accordance with the routing of the lines presented in
isometric drawings of the figures 3 and 4, was simulated with the computer program
CAESAR II [4], taking into consideration the load cases described in item 3.2 and performing
the stress analysis of piping according to:
The Primary Circuit Piping – nuclear safety class 2 (ASME III, subsection NC [5]);
The Secondary Circuit Piping – non- nuclear (ASME B31.1 [6]).
The maximum stress values calculated in the simulation do not exceed the limits set by
ASME code. These results were obtained considering the:
primary stresses, defined by equation (8) [5] for the Primary Circuit Nuclear Piping and
equation (11B) [6] for the Secondary Circuit Non-Nuclear Piping;
secondary stresses, defined by equation (10) [5] for the Primary Circuit Nuclear Piping
and equation (13A) [6] for the Secondary Circuit Non-Nuclear Piping.
The 3 highest values of stress, obtained from the computing simulation, in all load cases for
both conditions, are presented in the table 8.
Table 8: Stress Results Summary
Condition Equation Node stress (N/mm
2)
Ratio Nodal Allowable
Primary
Circuit
Design
1000 70,7 0,34
8 898 63,4 205,5 0,31
780 47,4 0,23
Thermal
Expansion
340 115,0 0,56
10 129 101,0 205,5 0,49
99 83,0 0,41
Secondary
Circuit
Design
3010 79,1 0,70
11B 3410 50,3 112,5 0,45
3480 42,1 0,37
Thermal
Expansion
2160 79,4 0,47
13A 2470 70,2 168,7 0,42
2100 53,3 0,32
The resulting stresses shown in table 8 are all bellow the limits of the ASME code.
INAC 2013, Recife, PE, Brazil.
5. CONCLUSIONS
The modifications in the IEA-R1 Primary Circuit Piping caused by:
“duplication of primary coolant system to redundancy”;
“installation of decay tank”;
“replacement of the heat exchanger”
The modifications in the IEA-R1 Secondary Circuit Piping caused by:
“installation of a new cooling tower”
“withdrawal of check valves”
“Replacement of secondary pumps”
This paper is the outcome of the structural analysis of the IEA-R1 coupled Primary and
Secondary Circuits piping after these modifications and changes in the layout. The structural
analysis developed in this paper is valid and can be considered correct, if the piping systems
are according to the following:
The piping, valves, tees, flanges and reductions must be without defects, it means: there is
no denting, no pitting, no misalignments, no out of roundness and nor any corroded points
on external or internal surface of the piping or fittings;
The attachments, welded or threaded, that connects piping to piping, piping to equipment
and piping to fittings (valves, tees, flanges, reductions) must be without defects, it means:
there is no denting, no pitting, no misalignments, no out of roundness and nor any
corroded points on external or internal surface of the fittings;
For the analysis purpose, the primary and secondary circuit of the reactor IEA-R1 was
modeled as a whole, encompassing piping and equipment coupled together. The Primary
Circuit Piping was classified as Nuclear Safety Class 2 (ASME III - subsection NC [5]) and
the Secondary Circuit Piping was classified as non nuclear (ASME - B31.1 [6]). Piping
calculation model was developed and the stress analysis was performed with the program
CAESAR II [4] considering the load cases presented below:
“DW” - Dead Weight; “PD” - Design Pressure; “Th1/ Th2/ Th3” - Thermal Expansion.
The values of the stresses calculated to the primary and secondary circuit piping do not
exceed the limits set by ASME code.
To sum up, the modifications made in the piping routing of the primary circuit for the
replacement of the heat exchanger, duplication of primary coolant system, installation of
decay tank and installation of a new cooling tower, replacement of secondary pumps and
withdrawal of check valves in the secondary circuit are guaranteed.
INAC 2013, Recife, PE, Brazil.
ACKNOWLEDGMENTS
We acknowledge the help of colleagues in doing this work, particularly the IEA-R1 staff,
which have facilitated the documents assessment and/or guided us into internal areas of the
IEA-R1 reactor building.
REFERENCES
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management”, AIEA – Projeto Arcal LXVIII “Seguridad de Reactores de Investigacion”
2002, Lima - Peru
4. CAESAR II, Structural Analysis Guide & Theory Reference Manual
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Components”, ASME, 2007.
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Piping B31, 2007.
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