Evaluation of the Potential for Containment Bypass due to Steam Generator Tube Rupture in VVER-1000/V320 Reactor during Extended SBO sequence using SCDAP/RELAP5 code

 a 
wide-scope PRA and the PRA model 
represents the set of accident sequences 
following the initial events (IEs) up to the end 
state of each sequence. The reports indicated 
that the most significant release category for 
offsite consequences is containment failure as a 
result of an IE with a leak from the primary to 
the secondary circuit [1]. In this aspect, steam 
generator tubes comprise a majority of the 
reactor coolant system pressure boundary and 
failure of steam generator tubes prior to the 
failure of one of other components (hot leg 
piping, pressurizer surge-line piping, and the 
reactor vessel) leads to discharge of some 
fission products into the steam generator 
secondary system from where they may be 
discharged to the environment through the 
pressure-relief valves. This sequence is 
potentially more risk-significant since it 
involves a containment bypass scenario. The 
relative timing of these structural failures 
therefore affects the event sequence and 
whether the containment is bypassed. 
This paper present thermal-hydraulic 
evaluations of VVER-1000/V-320 during 
extended SBO event using the 
SCDAP/RELAP5 systems analysis code. In 
general, the design features of VVER-1000/V-
338 are similar to the standard VVER-1000/V-
320, excepted of the main circulation loops 
EVALUATION OF THE POTENTIAL FOR CONTAINMENT BYPASS DUE TO STEAM 
10 
configuration which is equipped with main 
gate valves on the cold and hot legs (CLs and 
HLs) [2]. The potential for “Containment 
Bypass” is associated with natural circulation 
of superheated steam inside the piping loop 
that might induce creep rupture tube failure. 
SCDAP/RELAP5 predictions provide key fluid 
and components condition for estimation of the 
SG tube failure margins. 
II. DESCRIPTIONS OF ACCIDENT 
SCENARIO AND PLANT BEHAVIOR 
The low-probability SBO base case 
accident event scenario results in a severe 
accident because none of the systems that 
normally provide core cooling are assumed to 
be operable nor is any alternate equipment 
(e.g., security-related mitigation methods) 
assumed to be available [4]. The accident event 
is initiated by a loss of off-site alternating 
current (AC) power, which immediately results 
in reactor and turbine trips and the coast-down 
of the four reactor coolant pumps (RCPs). 
Also, assumptions are made as follows: 
 The diesel-electric generators fail to 
start and all AC power sources are lost; 
 Letdown flow is isolated and the 
pressurizer level control and RCP seal injection 
functions of the charging system are lost; 
 The high-pressure and low-pressure 
safety injection systems are unavailable as a 
result of the AC power loss; 
 The accumulator systems (four HAs) 
are available for injecting coolant into the cold 
legs should the RCS pressure fall below the 
initial accumulator pressure, 5.9 MPa; 
 The main feedwater (MFW) flow stops 
and the motor-driven auxiliary feedwater 
(MDAFW) system is unavailable as a result of 
the AC power loss; 
 The turbine-driven auxiliary feedwater 
(TDAFW) system is assumed to independently 
fail, so no MFW or auxiliary feedwater (AFW) 
system is available; 
 A station battery life of four hours is 
assumed; after that time all automatic and 
operator control of the pressurizer PORVs and 
SG secondary system PORVs is lost; 
During the initial portion of the accident 
scenario, buoyancy-driven coolant-loop natural 
circulation carries hot water from the core 
through the SGs, transferring heat to the SG 
secondary water inventory. The SG water 
inventory is boiled and the steam is released 
through the SG PORVs. The secondary water 
inventory declines and is eventually fully 
depleted since the MFW and AFW systems are 
not operative. After that time, the core decay 
power heats and swells the RCS water, 
increasing its temperature and pressure. The 
basic physical processes during this regard the 
transport of hot steam from the core outward 
into the other regions of the reactor vessel 
(RV) and coolant loops. Two coolant loop 
natural circulation flow patterns that may be 
encountered subsequent to the uncovering and 
heat-up of the reactor core, based upon whether 
or not the loop seals (the cold leg piping 
connecting the outlets of the SGs to the inlets 
of the RCPs) remain liquid-plugged. 
If liquid is cleared from a loop seal 
(along with the region of the RV lower plenum 
that extends above the bottom of the core 
barrel, the "downcomer skirt"), hot steam is 
transported from the core through the HL, SG 
tubes and cold legs in the normal direction of 
flow (i.e., that seen during plant operation). 
This flow pattern transports the hot steam 
directly (without benefit of mixing) through all 
of the SG tubes, leading to SG tube failure 
NGUYEN VAN THAI et al. 
11 
prior to HL or pressurizer surge line failure. 
However, if a loop seal remains liquid-
plugged, the more complex flow pattern 
develops instead. Hot steam is transported 
through the upper portion of the HL cross 
section to the SG inlet plenum, where it is 
mixed with cooler steam emanating from 
circulations set up within the SGs, with some 
of the tubes flowing in the normal direction 
and the remaining tubes flowing in the reverse 
direction. The mixing process within the SG 
inlet plenum determines the temperatures of 
the steam entering the SG tubes and the steam 
that is returned to the RV through the lower 
portion of the HL cross section. Fluid mixing 
in the SG inlet plenum buffers the entry of hot 
steam into the SG tubes, thus delaying SG tube 
failure and making it more likely that some 
other component (HL, pressurizer surge line or 
RV) will be the first to fail. 
The issues of primary interest for 
containment bypass are: (1) do the loop seals in 
all coolant loops remain liquid plugged and if 
so (2) does the fluid mixing in the SG inlet 
plenum sufficiently slow the SG tube heat-up 
process so that the HL, pressurizer surge line 
or RV will fail prior to a SG tube? [4]. The 
analyses in this paper address these issues. 
Fig. 1. RCS Layout of VVER-100/V-320 [3] 
471
472
473
478
479
461
462
463
464
465
466
474
475
476
171
172
173
178
179
161
162
163
164
165
166
174
175
176
15-1714-16
18-20 19-21
22-24 23-25
26 28 29 27
1
2
3
4
5
6
7
8
9
30 32 33 31 34
4
3
2
1
35
1 1
37-4036
6
41-44
6
7
8 45-48 51-52 63-64
3-4
54
56
7-8
6
5
4
3
2
1
6
5
4
3
2
1
10-129-11
61-62
1-2
5-6
425
424
423
422
421
400
1 2 3 4
504
506
502
402
410
411
412
413
414
415
416 436
435
434
433
432
431
430
11
444 441442
9
440
125
124
123
122
121
100
1234
102
110
111
112
113
114
115
116136
135
134
133
132
131
130
11144141 142
9
140
167
168
169
Separator
Volume above 
perforated sheets
Steam 
dome
177
Volume between 
tube bundles and 
perforated sheets
150
152
Volume between 
tube bundles and 
perforated sheets
450
452
467
468
469
Separator
Volume above 
perforated sheets
Steam 
dome
477
480
482484
490
492
180
182 184
190
192
186 486286 386
194
196
TURBINE
MSIV
187
MSIV
287
MSIV
387
MSIV
487
1
508
511
PORV
49-50
ACC ACC ACCACC
JNG50
086
JNG60
087
JNG70
088
JNG80
089
REG 
Valve
185
REG 
Valve
285
REG 
Valve
385
REG 
Valve
485
BRU-A1
BRU-In1
BRU-A1
BRU-In1
191
189 489
491
Fig. 2. Nodalization scheme of VVER-1000/V-320 Plant 
EVALUATION OF THE POTENTIAL FOR CONTAINMENT BYPASS DUE TO STEAM 
12 
III. PLANT NODALIZATION AND 
PARAMETERS SETUP WITH 
SCDAP/RELAP5 
The SCDAP/RELAP5 computer code 
calculates the overall RCS thermal-hydraulic 
response for severe accident situations that 
include core damage progression and RV heat-
up and damage. The computer code is the result 
of a merging of the RELAP5 and SCDAP 
computer codes. Prediction of structural failure 
are made based on the structure configuration, 
its material properties and the fluid conditions 
that are locally present. The code also includes 
models for calculating the creepture failure of 
structural components and these are used to 
predict failure time for the hot legs, pressurizer 
surge line, and SG tubes [5]. 
The SCDAP/RELAP5 VVER-
1000/V320 plant model represents the fluid 
volumes and structures in the core, RV and 
primary and secondary coolant system regions 
in the plant as shown in Fig 1. The nodalization 
diagrams for the final SCDAP/RELAP5 
VVER-1000/V-320 four-loop plant model are 
provided in Fig. 2. 
In this work, creep rupture failure 
calculations are performed for the hot average SG 
tubes in which pressure and temperature 
conditions of fluid and heat structures in four SGs 
are used as inputs for the analyses. A parameter 
that measure creep damage is calculated each time 
step for each structure being monitored for creep 
rupture by following equation: 
 tt
t
tDttD
r
cc
 (1) 
Where tDc is creep damage at 
problem time t, t is time step at current 
problem time and ttr is time required for the 
structure to fail by creep rupture at current state 
of temperature and stress. If the value of Dc is 
zero, the structure has not experienced any 
creep damage. If the value is one, the structure 
has failed due to creep damage. Two different 
theories (Larson Miller and Manson-Haferd) 
are applied in SCDAP/RELAP5 which are 
dependent upon the structural materials and the 
range of stress and summarized in Table I. 
Table I. Equations for the time to creep rupture [5] 
Since SG tubes are made from heat-
resistant steel (08X18H10T) which is similar to 
heat-resistant Stainless Steel, equations for 
creep rupture time of 316 Stainless Steel in 
Table i are selected in this study. Estimated 
stress in SG tubes (cylindrical shaped 
structure) is calculated as follows: 
 io
ooii
rr
rPrP
)(
 (2) 
Fig. 3. Model ofSG ubes 
IV. RESULTS AND DISCUSSION 
A. Plant behaviors 
The SCDAP/RELAP5 plant model was 
run to a steady solution. The plant model using 
the nodalization scheme to establish full-power 
steady-state conditions from which the SBO 
transient accident sequence is initiated 
(Figures.4&5). 
NGUYEN VAN THAI et al. 
13 
Fig. 4. Steady-state conditions (Pressure) at primary 
and secondary loops 
Fig. 5. Steady-state conditions (Mass Flow Rate 
and Temperature) at primary and secondary loops 
The SBO base case event sequence was 
simulated with SCDAP/RELAP5, starting from 
time 250s when the loss of off-site power 
occurs. Figure 5 showed the transient condition 
of reactor power and pressure level in PRZ. 
After a short period when the PRZ 
pressure initially falls and rises slightly due to 
the effects of the reactor and turbine trips, the 
PRZ pressure declines in response to the 
cooling provided by heat removed to the SGs. 
The RCS fluid mass lost through the 
pressurizer PORVs and SRVs and through the 
RCP shaft seal leakage paths depletes the 
RCS inventory. 
Fig. 6. Transient condition of reactor power and 
pressure in PRZ 
Fig. 7. Transient condition of mass flow rate 
through the primary loops and SG levels 
Buoyancy-driven coolant-loop natural 
circulation carries hot water from the core 
through the SGs, transferring heat to the SG 
secondary water inventory until the SG heat 
sink is lost around 11000s (Fig. 7). It can be 
seen that after the SG heat sink is lost, the 
pressure the cooling afforded by system heat 
loss to containment and RCP shaft seal leak 
flow is insufficient to remove the RCS heat 
load, causing the RCS and PRZ pressure to 
increase (Fig. 6). After SG dry-out, the RCS 
pressure increase is limited by multiple cycling 
of the PORVs and by two cycles of the SRVs 
during the period with the most-challenging 
RCS pressurization conditions. This challenge 
EVALUATION OF THE POTENTIAL FOR CONTAINMENT BYPASS DUE TO STEAM 
14 
is presented when the increasing temperatures 
cause the RCS fluid to swell, completely filling 
the pressurizer with water. The fuel 
temperature increases after the heat sink is lost 
and the onset of fuel rod oxidation as well as 
fuel rod cladding rupture occurred as an 
unvoidable consequence (Fig. 8). 
Fig. 8. Temperature of fuel pins and hot legs 
B. Creep Rupture Analysis of SG Tubes 
Creep rupture model allows one to 
specify a "stress multiplier" in following 
meaning: a multiplier of 1.0 provides a creep 
rupture failure prediction based on no 
degradation of the structural strength of the 
material and multipliers greater than 1.0 
represent degraded structural strengths. In this 
study, a stress multiplier from 1.0 to 5.0 (in 
increments of 0.1) are used for SG tube 
rupture prediction to investigate the spectrum 
of material strengths, from undergraded to 
highly-degraded. 
Typical results on Larson-Miller creep 
rupture damage index for the hottest SG tubes 
is presented in Fig. 9. It was found that the 
average-tube predicted failure margin is 2.3 
and the hottest-tube predicted failure margin is 
1.7. Also, SG tube failure margin are 
insensitive to variations in the fuel damage 
progression behavior primarily because these 
occur after the time when SG tube structural 
failures is experienced. 
Fig. 9. Structure temperature and creep 
damage index 
V. CONCLUSIONS 
The paper presents preliminary 
assessments the VVER-1000/V-320 plant 
behavior during an SBO event with the 
SCDAP/RELAP5 for key fluid and 
components condition and for the SG tube 
failure margins. Further analyses will be 
performed to evaluate the potential for 
“Containment Bypass” in VVER-1000/V-320 
reactor during extended Station Blackout 
(SBO) scenarios. 
ACKNOWLEGEMENTS 
This work was supported by the 
Ministry of Science and Technology (MOST) 
R&D Project with contract number 
05/HĐ/ĐTCB signed on 05/01/2018. 
.REFERENCES 
[1]. Federal Nuclear and Radiation Safety 
Authority of the Russian Federation, Kalinin 
VVER-1000 Nuclear Power Station Unit 1 
PRA (Beta Project), NUREG/IA-0212, 
published by US NRC, 1992. 
NGUYEN VAN THAI et al. 
15 
[2]. Design, Safety Technology and Operability 
Features of Advanced VVERs, OKB 
“GIDROPRESS’’, 2011. 
[3]. S., Pylev. Assessment Study of 
RELAP5/MOD3.2 Based on the Kalinin NPP 
Unit-1 Stop of Feedwater Supply to the Steam 
Generator No. 4, NUREG/IA-0167, published 
by US NRC, 1999. 
[4]. C.D., Fletcher et al. SCDAP/RELAP5 Thermal-
Hydraulic Evaluations of the Potential for 
Containment Bypass During Extended Station 
Blackout Severe Accident Sequences in a 
Westinghouse Four-Loop PWR. NUREG/CR-
6995, published by US NRC, 2010. 
[5]. C. Allison et al., “SCDAP/RELAP5/MOD2 
Code Manual”, Volume: I, II and III, 
NUREG/CR-5237, EGG-2555, June 1989.