VGB POWERTECH 7 (2021) - International Journal for Generation and Storage of Electricity and Heat

Verification of SPACE code based on an MSGTR experiment at the ATLAS-PAFS facility VGB PowerTech 7 l 2021
ty analysis codes supplied by foreign vendors
such as Westinghouse and Combustion
Engineering have been used. The
Ministry Of Trade, Industry and Energy
(MOTIE) in Korea launched the ‘Nu-Tech
2012’ project to improve the competitiveness
of the Korean nuclear industry in
2006, and the Korean nuclear industry developed
the SPACE (Safety and Performance
Analysis CodE for nuclear power
plants) code [6]. This code was approved
by the Korean NSSC in 2017, and it will replace
outdated vendor-supplied codes and
will be used for safety analyses of operating
nuclear power plants in Korea as well as
the design of advanced reactors. While the
SPACE code will mainly be used in safety
analyses, the code with best-estimate capabilities
will be able to cover performance
analysis as well. The programming language
for the SPACE code is C++, and
this code adopts the advanced physical
modeling of two-phase flows, namely
two-fluid three-field models which comprise
gas, continuous liquid, and droplet
fields.
According to the General Safety Requirement
(GSR) of IAEA, any calculation method
and computer codes used in safety analysis
must undergo verification and validation
[7]. Further, verification calculations
of system tests or integral tests are used to
validate the general consistency of the revision
[8]. Therefore, a verification calculation
based on integral effect experiments is
needed to improve the reliability of the prediction
results of the SPACE code. In particular,
the multiple failure accident condition
is the new safety design criteria, so the
verification process should be performed.
To this end, the first objective of this study
is to verify a multiple failure accident calculation
capability of the SPACE code,
while the second objective is to confirm the
transient phenomena of MSGTR and the
cooling effect of PAFS during MSGTR with
respect to the first objective, as mentioned
above. The heat loss phenomenon is a
measure of the total heat transfer of heat
from either conduction, convection, radiation,
or any combination of these. Newton’s
law of cooling states that the rate of
heat loss of an object is directly proportional
to the difference in the temperature between
the object and its surroundings. Under
the experimental conditions of high
temperature and high pressure, the heat
loss is particularly likely to increase because
of the temperature difference between
the experiment component and the
surrounding atmosphere. This physical
phenomenon can affect the heat transfer
experiment, and it plays an important role
in the performance of the system. The heat
loss is a function of area in accordance with
the convective heat transfer equation. According
to the design document, the AT-
LAS facility has a relatively large surface
area to volume ratio which is in accordance
with the design characteristic [9]. For this
reason, additional work to confirm the heat
loss effect on the ATLAS-PAFS facility was
also conducted to confirm the heat loss effect
on the system behavior of the integral
test facility.
1.2 A brief description of ATLAS-PAFS
facility
As mentioned previously, KAERI has been
operating an integral effect test facility,
‘ATLAS’, for the transient and accident simulation
of a Pressurized Water Reactor
(PWR) [10]. The reference plant of ATLAS
is the APR1400, which has been developed
by the Korean nuclear industry. ATLAS has
the same two-loop features as the APR1400,
and it is designed using the scaling method
to simulate various test scenarios as realistically
as possible [9, 10]. This test facility
also includes design features of the
OPR1000, which is Korean standard NPP,
such as a cold-leg injection mode for safety
injection and a low pressure safety injection
mode.
As mentioned above, the PAFS is one of the
advanced passive safety systems adopted
in the APR+, and an experimental program
is currently underway at KAERI to
validate the cooling and operational performance
of the PAFS [11]. The main objective
of the ATLAS-PAFS integral effect
test is to investigate the thermal hydraulic
behavior in the primary and secondary systems
of the ATLAS during a transient at
which PAFS is actuated. The PAFS facility is
described in further detail below.
2. Description of ATLAS-PAFS
model for MSGTR scenario
using SPACE code
2.1 Experiment scenario
To validate the prediction capability of the
SPACE code, the experimental information
provided by KAERI was utilized [4]. The
target scenario for the experiment is a
MSGTR with a PAFS operation occurrence
and asymmetric cooling. To initiate the
MSGTR transient, first, the break valve at
the SGTR simulation pipe line is opened.
By opening the break valve, the primary
system inventory was discharged from the
hot side of the lower plenum to the upper
location of the steam generator secondary
hot side. Next, the primary system began to
be depressurized and the secondary side
water level of the steam generator increased.
At the same time as the HSGL signal
occurrence, reactor trip occurred, and
the core power started to decrease following
the decay curve. For the transient calculation,
the decay power curve was considered
along with the tables for time versus
power. The main feedwater isolation
valves (MFIVs) and the main steam isolation
valves (MSIVs) for two steam generators
were closed after delay times. The
main steam safety valves (MSSVs) on the
steam line opened due to the pressure increase
of SG-1, and these valves were kept
in cyclic operation of opening and closing
to protect the primary and secondary systems
from over-pressurization. The accident
caused the depressurization of RCS
and resulted in a Low PZR Pressure (LPP)
signal. Further, the Safety Injection Pumps
(SIPs) began after delay times. It was assumed
that only one safety injection pump
per train was operated for the experiment
scenario. In accordance with this assumption,
SIP-1 and SIP-3 were available. The
injection flow rate was applied using pressure
- mass flow curve based on experiment
data. To simulate an accident management
measure based on the cooling
performance of PAFS during a MSGTR, the
PAFS was supplied to an intact SG-2 instead
of auxiliary feedwater after the water
level of the SG-2 fell below the PAFS operation
set point due to the decay power. It
was also assumed that the auxiliary feed
water system would not work for the assessment
of PAFS cooling capability. After
the initiation of the PAFS, the decay heat
was removed from the RCS by the natural
convection of the PAFS. Finally, the whole
system was cooled down in a stable manner
with the successful operation of SIPs,
MSSVs, and PAFS.
2.2 Brief overview of model
information
For the experimental simulation, the AT-
LAS-PAFS test facility was modeled using
SPACE code as shown in F i g u r e 1 . In the
calculation, the decay power was imposed
in accordance with the experiment, and
the operation logics of the safety systems,
such as safety injection and PAFS, were reflected.
The geometrical and material information
of the components and pipe
lines in the ATLAS-PAFS test facility was
also reflected [9, 12]. The reactor vessel
was separated to simulate the core, bypass
flow, reactor vessel lower plenum, and the
reactor vessel upper head. The core model
included the top and bottom inactive core
regions, average channel, and hot channel.
The safety injection system had four independent
trains and a direct vessel injection
(DVI) mode. Each train of the safety injection
system consisted of safety injection
pumps (SIPs). Injection lines could be
aligned to either reactor vessel down-comer
for DVI injection. The pressurizer was
modeled as a single component with 10
vertical nodes. The lower part was connected
to hot leg through a surge line separated
into five nodes. The hot legs and cold
legs were modeled with four cells each,
while the intermediate legs were modeled
with five nodes. The steam generator included
five nodes for the evaporator and
two nodes for the economizer. The main
steam safety valves were modeled into
three separate groups; each group was operated
on a different set point of pressure
in the steam generator dome.
78