The safety evaluation which is described in the previous sections is
a deterministic safety evaluation. In the evaluation a number of
representative events are postulated so as to include various
probable events. The representative events are analyzed to ascertain
the acceptance criteria are met
On the contrary, probabilistic safety assessment (PSA) is a
comprehensive method which assesses the occurrence frequency of
abnormal events or accidents at NPPs (i.e. initiating
events), combines failure probabilities of individual
systems and components to mitigate the events and then assesses
progression of the events and degrees of their consequences as a
whole. The PSA approach was established after the reactor safety
study WASH-1400 (Rasmussen Report) was published in 1975. The PSA
standards for the assessment procedures, etc. have been published in
Japan and PSAs have been conducted by Japanese research institutes
and industries. PSAs have included core
damage frequency (level 1 PSA), loss of reactor
containment function (level 2 PSA), etc.
The core damage frequency and loss of containment function frequency
obtained by PSAs are expressed by the units of “CHO
c'/ry”
as the probabilities that the nuclear reactor facility is lead to
such a state that its core is damaged or the reactor containment
loses the function in one year
of operation of the nuclear reactor plant (ry: reactor year) .
Level 1 and level 2 PSAs are described below, taking a Japanese
4-loop plant as an example.
In the level 1 PSA during power operation of a plant, initiating
events which may render the reactor into an abnormal state are first
identified and safety criteria for attaining reactor safe shutdown
are defined. Then event-trees are prepared which describe the
progression of events. System models are developed by fault-trees,
etc. for the elements (headings) of the event-trees to compute
subordinating failures and analyze human reliability. After
preparation of the database necessary to quantify accident
sequences, core damage frequencies are assessed.
Besides, based on the assessment results of core damage frequency,
and focusing mainly on failed safety functions to grasp the
characteristics of the plant safety aspects, the accident sequences
are quantified and classified into 7 categories: “Loss of ECCS
Recirculation Function,” “Loss of ECCS Injection Function,”
“Loss of Isolation of Leaking Portions,” “Loss of Heat Removal
Function from Secondary Systems'1,
“Loss of Supporting Functions of Safety Functions,” “Loss of
Reactor Shutdown Function,” and “Loss of Containment Heat
Removal Function.”
Assessment results of level 1 PSA during power operation are shown
in Figure 8.5.1 (1).
The categories of “Loss of ECCS Recirculation Function” and
“Loss of ECCS Injection Function,” contribute significantly to
core damage frequency with comparatively high fractions.
In level 2 PSA at power operation of a plant, from the results of
level 1 PSA, accident sequences resulting in core damage are grouped
at first as plant damage states. Then containment eventtrees
are developed by elaborating the physical phenomena generated inside
the containment, measures for prevention and mitigation of the
event, and by selecting the elements (headings) of containment
event-trees and the modes that threaten the integrity
of the containment.
At last the progressions of accident sequences, grouped by
similarities of the event progression in the reactor vessel and
inside the containment, are assessed, and the frequency of loss of
containment integrity is quantified in relation with the core damage
probabilities of the accident sequences.
Besides in order to grasp the characteristics of the reactor plant
safety from the results of frequencies of loss of containment
function , the quantified accident sequences are classified into 9
categories, focusing mainly on the containment failure modes:
“Overpressure due to Steam (Decay Heat),” “Loss of Isolation
of Leaking Portions,” “Concrete erosion,” “Loss of
Containment Isolation Function,” “Penetration Over-temperature,”
“Steam Explosion,” “Burning of Flammable Gas at
High Concentration,” “Direct Heating of Containment
Atmosphere,” and “Direct Contact to Containment”
The results of level 2 PSA at power operation are shown in Figure
8.5.1 (2).
NSRA,
Japan
8
—
20Probabilistic Safety Assessment (psa) for pwr Plants
Outline of Probabilistic Safety Assessment
Fraction
of Contribution for Containment Integrity Categories
Figure
8.5.1 (2) Results of PSA for containment vessel integrity (Level-2
PSA; 4-loop plant)
Figure
8.5.1
(1)
Results of psa for core integrity (Level-2
PSA; 4-loop plant)
Fraction
of Contribution for Core Integrity Categories
Chapter
8 Safely Evaluation of PWR Plants
The categories of “Overpressure due to Steam (Decay Heat)”
and “Loss of Isolation of Leaking Portions,” etc. contribute
largely to the frequency of loss of containment function.
Example Uses of PSA in
Japan
(1) Implication of
accident management
NPPs are provided with safety systems against very rare
accidents. The events postulated in the design are called design
basis events. If a beyond- design-basis event occurs which has a
possibility for large core damage, measures called Accident
Management (AM) are provided to prevent the core damage from
being enlarged or to mitigate the consequences of the severe
conditions; using the safety margin in the current safety design,
functions expected to be effective in addition to the principal
safety functions prepared in safety design and new equipment are
prepared for managing such events.
Japanese electric utilities have reflected on the problems
experienced inside and outside Japan and they positively
investigated AM measures from the standpoint of voluntary safety
assurance activity after the Three
Mile Island Nuclear Power
Plant No.2 accident in
1979.Further, they have
been promoting measures to be taken in response to occurrence of
beyond-design-basis
events by effectively utilizing systems in the NPPs.
Moreover, Japanese NPP operators have conducted PSAs for abnormal
events caused by systems failures during power operation of the
plants (Internal Events), and developed AM measures to improve
safety based on the PSA knowledge and on new knowledge about the
physical phenomena in severe accidents. By 2002 AM measures had
been completed for all operating plants in Japan. Table 8.5.1
summarizes AM measures prepared for PWR plants.
From the PSAs conducted before and after the preparation of the
AM measures, it was ascertained that the core damage frequency
and loss of containment function frequency are significantly
reduced and the measures are effective to enhance reactor plant
safety further. Figure 8.5.1(1)
and Figure 8.5.1(2)
shows examples of PSA
results obtained before and after the implementation of AM
measures.
Table
8.5.1 Accident management measures (4-loop plant) |
Implicated Accident Management Measures |
(1) Function of Reactor Shutdown |
(1) Diversity of Emergency Secondary Cooling |
(2) Core Cooling Function |
(D Use of Turbine Bypass System c' Low Pressure Injection by Secondary Forced Cooling "i Low Pressure Recirculation by Secondary Forced Cooling Secondary Forced Sump Water Cooling L Alternative Steam Discharge J (D Alternative Recirculation (3) Natural Convection Cooling inside Containment Vessel ® Alternative Auxiliary Component Cooling (5) Cool-down and Recirculation |
(3)Confinement Function of Radioactive Material |
@ Primary Forced Depressurization |
(4) Supporting Function of Safety Function |
© Alternative Auxiliary Component (2) Electrical Power Tie between Units |
NSRA,
Japan
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Chapter
8 Safety Evaluation of PWR Plants
(2) PSA at the time of
periodical safety review
In the periodical safety review (PSR) **
5 program, PSAs during power operation and in the
shutdown state of NPPs were performed to ascertain the safety
levels. The results of the PSA studies were well under the goals
(less than lOVry for
operating plants) stated in the “Basic Safety Principles for
Nuclear Power Plants (INSAG-12), 1999” (International Atomic
Energy Agency IAEA) and
the performance goals required in the “Performance Goals for Light
Water Nuclear Power Reactor Facilities” (Nuclear Safety
Commission, Special Committee on Safety Goals in Japan) issued in
2006 (core damage frequency, about lOVry;
loss of containment function frequency, about lOVry;
both should be satisfied at the same time). By these assessments the
safety levels of the NPPs are ascertained to be sufficiently
ensured.
Containment Recirculation Sump
r
V Area of System
i
| Modification
Containment
SprayCooler
T—
r
Reactor
Containment
I
Spray System.
Containment
Spray Pump
Residual Heat Residual Heat
Removal Cooler Ph
mo
(Residual Heat Removal System)
< V
(Failure of Recirculation)
In “Alternative
Recirculation”
The ECCS cools the reactor in
the event of a reactor coolant piping break, etc. by injecting
aqueous boric acid (stored in an outside tank) and re-circulating
the water collected in the recirculation sump (water of the ECCS and
primary coolant that are came from the piping break, etc.) after the
tank is emptied. Alternative recirculation is a preparatory
provision to inject core cooling water through reactor containment
spray system by connecting residual heat removal system to the
containment spray system when the function of the preferred
redundant recirculation
operation is lost.
Figure
8.5.2 Conceptual figure of “alternative recirculation”
(4-loop Plant)
5'
Activities of periodic safety review (within every 10 years) of each
nuclear reactor are carried out by the operator to assess"Status
of Safety and Maintenance Activities for Nuclear Reactor Facilities”
and “Status of Reflection of Recent Technical Knowledge to Safety
and Maintenance Activities for Nuclear Reactor Facilities”
in Accordance with the National Law: TRules
on Establishment and Operation, etc. of Power Generating Nuclear
Reactors’.
8-23
NSRA,
Japan
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