atw - International Journal for Nuclear Power | 06.2021

atw Vol. 66 (2021) | Issue 6 ı November
cause an initial event and equipment
that should be used to mitigate safety
shutdown such as engineering safety
feature when a seismic-induced initial
event occurs. Equipment that may
cause an initial event should be
included conservatively without a
screening process, as all of the
initial events may cause direct core
damage. However, this equipment
may be excluded if the frequency of
occurrence is calculated to be below
1.00E-07/yr, which corresponds to
the typical initial event screening
criteria [12]. Table 9 shows equipment
that can initiate seismic event
in general light-water reactor nuclear
power plant.
In general, seismic events can
occur in seismic-induced loss of
coolant accident, loss of power, loss of
control, loss of ultimate heat sink,
main steam line break, and anticipated
transient without scram.
Next, when the seismic-induced
initiating event occurs, an analysis
of the equipment considered to
mitigate the event is performed. Like
all external event analysis, SPSA
basically uses the event tree and fault
Seismic initiating event
Loss of
Coolant Accident
(LOCA)
Loss of Power
(LOP)
Loss of Control
(LOC)
Loss of Heat Sink
(LOHS)
Large
Medium
Small
Small-Small
Loss of offsite power
Station black out
Loss of KV
Loss of DC power
Loss of AC power
Main control room
Component cooling
water
Ultimate heat sink
Main steam line break (MSLB)
Anticipate transient without scram (ATWS)
Building failure
| Tab. 9.
Equipment for seismic induced initiating event.
CASE
| Tab. 10.
Sensitivity analysis result for importance analysis of internal events PSA.
tree used in the internal event PSA.
Here, non-safety and non-seismic
equipment that cannot be used conservatively
is excluded from the
model. However, although seismicinduced
initiating events are different
from internal events, the primary
heat removal, which is essential for
mitigation of the accident, must be
performed by the same equipment
and procedure. Therefore, in the
ASME Standard, which is considered
the standard of the PSA, equipment
corresponding to the standard can be
listed based on FV (Fussell-Vesely)
importance 0.005 higher and RAW
(Risk Achievement Worth) value of
two or higher, which are the criteria
that can be used to classify significant
basic events. To confirm the application
of importance value, a sensitivity
analysis was performed on the
reference nuclear power plant. In
Table 10, we show the difference between
the results of the existing SPSA
and CDF when only the equipment
that was evaluated as important in
internal events was modeled.
As a result, it is confirmed that the
results of the sensitivity analysis
Equipment
Pressurizer, Steam generator, RCP, RCS
RCP seal
Impulse lines
EDG
DC panel, Inverter, Battery, Charger
AC Panel
Main control board, MCR HVAC
Pump, Heat exchanger, Piping
Pump, Heat exchanger,
Intake structure
Control rods
Containment building
Service building
Turbine building
# of basic event
for SPSA model
% of
baseline CDF
Base model 1,798 100.0 %
Only FV important basic event 49 97.0 %
Only RAW important basic event 217 95.9 %
Intersection FV and RAW
important basic event
24 95.5 %
Union FV or RAW important basic event 242 97.6 %
showed no significant difference
between the case of modeling all
equipment and the case of modeling
only equipment that was evaluated as
important in internal events. When
242 items of equipment, 13 % of the
total of 1798, were modeled in the
SPSA model, the CDF was 97.6 % of
the baseline CDF, and even when
49 pieces of equipment classified as
important in basic events based on FV
are modeled, a result equivalent
to 97 % of the baseline CDF was
obtained. In addition, even when only
24 pieces of equipment were considered,
a value corresponding to
95.5 % of the CDF value of the base
model was derived, indicating that
the importance of other equipment
other than this was much lower
than expected. Therefore, it is
judged that the equipment selection
methodology based on the values of
the internal event FV and RAW is
much more rational than the
existing HCLPF-based method, and is
an efficient method that can
reflect the characteristics of the
SPSA model.
4.4 Summary of the proposed
equipment selection
methodology
The procedure is shown in Figure 3, a
flow chart of the method of selecting
equipment for seismic events over the
three steps described above.
First, based on the site seismic
hazard curve obtained as a result of
PSHA, the SSE value is checked,
which is the design standard of the
power plant, and probability value
exceeding SSE value is checked.
After that, the region of ​the site is
determined by checking the SSE
reoccurrence period according to
the SSE excess probability value.
Through this value, the group of
equipment that should perform
fragility analysis is determined. Next,
based on the results of the PSA
importance of internal events, basic
events with an FV value of 0.005 or
more or RAW value of two or more,
are listed and described in terms of
the function of the system. However,
basic events related to non-seismic
equipment and human error are
excluded among the results of the
PSA importance of internal events.
Finally, equipment is derived
through cross-examination between
the ‐equipment required for the
function of the mitigating equipment
derived from an internal event and the
equipment group subject to fragility
analysis.
ENVIRONMENT AND SAFETY 33
Environment and Safety
Equipment Selection Methodology of Seismic Probability Safety Assessment for Nuclear Power Plant ı Junghyun Ryu and Moosung Jae