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

VGB PowerTech 7 l 2021 Continuation of table 8
relatively low, and the probability of exceeding
0.2 g on the SSE basis corresponds
to 1.06E-04/yr, corresponding to the Region
C medium risk region proposed in this
paper. In terms of the likelihood of earthquake
occurrence, it is a value that is approximately
54 % of the total power plants
in terms of seismic hazard considered by
F i g u r e 2 . Based on this, the plant needs
to review equipment groups I and II subject
to fragility analysis. Fragility analysis
groups I and II include general vulnerable
equipment including offsite power sources
and yard tanks, as well as general active
equipment. The next analysis is the plant
seismic response analysis using internal
events PSA results; 74 basic events were selected
with an FV value of 0.005 or more
and 253 basic events with a RAW value of
two or more. Among them, 30 pieces of
equipment that satisfy both FV and RAW
are considered, along with 297 basic
events. Excluding 20 human error basic
events and 26 non-seismic basic events,
251 basic events are considered. In addition,
except for valves, flow elements, flow
transmitters, radiation transmitters, dampers
and filters, which are inherently rugged
SSCs, the total 136 basic events are derived.
The 136 basic events derived can be divided
into related systems and equipment
types to achieve the following 14 critical
system functions. Here, if we review the 14
important functions, we can confirm easily
the unique operating characteristics of the
reference plant.
––
Auxiliary feedwater (AF) supply
––
AF pump room cooling
––
Emergency power supply
––
Component cooling water supply
––
Diesel generator fuel supply
––
Diesel generator room cooling
––
Essential chilled water supply
––
High-pressure injection
––
Low-pressure injection
––
Plant control
––
Ultimate heat sink
––
Reactor containment cooling
––
Safety actuation signal
––
Ultimate heat sink pump room cooling
In the end, cross-examining 14 critical systems
and functions with fragility equipment
groups I and II, the results shown in
Ta b l e 11 are obtained.
The number of equipment items derived
through cross-examination is a total of 23,
which is a very small result considering the
overall equipment in nuclear power plants.
However, when reviewing the previously
analyzed SPSA results, it can be confirmed
that all devices are considered important in
the existing SPSA model except for the three
pieces of equipment that initiate seismicinduced
initiating events, so the methodology
proposed in this paper is very efficient.
It can be confirmed that it is reasonable.
6. Conclusion
In this study, a methodology of the equipment
selection for SPSA is proposed. The
single HCLPF screening criterion which
has been applied for SPSA reflects some of
the site-specific PSHA results but does not
reflect the plant design characteristics, so if
the single HCLPF screening criterion is applied
to the model based on this, an optimistic
evaluation can be made. The methodology
proposed in this paper has the following
advantages:
Safety aspects: Single HCLPF screening criteria
is not applied, and all equipment is
not reflected in the model using general
fragility data, so realistic and reasonable
SPSA results are expected.
Economics aspects: As the largest portion
of the required manpower for SPSA is the
detailed fragility analysis work, the methodology
proposed in this paper is economically
beneficial as the number of pieces of
equipment subject to SPSA decreases.
The equipment selection methodology proposed
in this paper requires analysis of all
three parts of SPSA, unlike the previously
proposed methodology. This means that
selecting equipment through consideration
of only one part, such as PSHA, may
lead to incorrect results. SPSA has its own
uncertainty, so if one factor affects several
steps, the uncertainty becomes very large.
Therefore, the analysis of equipment not
essential for SPSA can increase this uncertainty,
so it can be said that it is necessary
for a much more realistic analysis that is
not considered in advance.
In recent years, SPSA has evolved from a
single unit criterion evaluation to an evaluation
of multiple units operating at one
site. In this case, the number of equipment
items handled by the SPSA increases, resulting
in unnecessary model enlargement.
Therefore, even in such case, it is necessary
to select equipment that has a significant
simplification of the SPSA model and the
accuracy of the analysis of the quantitative
results, and minimize unnecessary analysis.
Therefore, it is expected that uncertainty
errors will be minimized.
Acknowledgments
This work was supported by the Nuclear
Safety Research Program (No. 2101052)
through the Korea Foundation Of Nuclear
Safety (KOFONS), Republic of Korea
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