Chapter
7 Safety Evaluation of BWR Plants
transients as initiating events are dominant. This comes from the
design differences that the ECCS of BWR-3 plants is composed of
three systems, but that of BWR-4/5 or ABWR plants has five or more
systems as a result of their enhanced design against LOCA
Furthermore, for the high pressure core cooling system that is
important during transients, BWR-3 plants are provided with two
trains of reactor isolation condensers (ICs) as a passive cooling
system in addition to the high pressure core injection system
(HPCI), but BWR-4/5 plants are provided with two trains of HPCI.
Therefore, for BWR-3 plants, the contribution is small due to
sequences with failures of high pressure injection and
depressurization (sequence whereby high pressure core cooling fails
during transients, depressurization of a pressure vessel also fails,
making water injection with the low pressure core cooling system
impossible). Since ABWR plants are provided with three high pressure
core cooling systems, the contribution of failures of high pressure
injection and depressurization is smaller compared with that of
BWR-4/5 plants, and the overall core damage frequency is decreased.
Relatively, the contribution of the loss-of-power sequence (station
blackout) becomes larger.
Moreover, the loss-of-power sequence is not dominant for BWR-3
plants with ICs provided. And the probability of failure to ensure
non-criticality (ATWS) in terms of core damage frequency is
sufficiently reduced for
all plant types.
The PSAs have been performed as voluntary evaluations by reactor
licensees together with periodic safety reviews (PSRs) conducted by
reactor licensees, in accordance with the "Rules for
Failure
to ensure
subcriticality
Failure
of
high
pressure
and low
pressure water
injection
Failure
of decay heat removal
Loss
of power
Failure
of high
Failure
of water
tLOCA
injection
pressure water
injection and
depressurization
BWR-3 BWR-4
Failure
to ensure
subcriticality
Failure
of high
pressure and low
pressure
water
injection
Failure
of water injection at LOCA
Loss
of power
BWR-5
Failure of
high
pressure and low
pressure water
injection
Failure of decay
heat removal
Failure
of decay
heat
removal
Failure of high
pressure
and low
pressure water
injection
Others
Loss
of power
ABWR
Failure of high
pressure water
injection and
depressurization
7
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Figure 7.5.1 Contribution of each sequence to the core damage frequency
Installation, Operation, etc. of Commercial Power Reactors"
established by the National Government, and it has been confirmed
that the results conform to the international and Japanese targets.
The Japanese targets are provided in the "Performance Target of
Light Water Nuclear Power Reactor Facilities; Performance Target
Corresponding to the Proposal of Safety Goals" (Safety
Objective Special Committee, Nuclear Safety Commission, 2006), which
describes that both the core damage and the containment failure
frequencies of approximately 10"
and 10 s per
reactor year should be satisfied, respectively. Moreover, not only
the PSAs during power operation but also PSAs during reactor
shutdown etc. have been studied, and the National Government,
academic societies, and industrial associations have been promoting
preparation of the related standards for performing PSAs together
with a method to utilize the risk information obtained from the
PSAs.
As mentioned above, the PSA during BWR operation is evaluated from
the viewpoint of a "significant core-damage event", this
viewpoint is also the.same
for the PSA during shutdown.
That is, in evaluation of the
significant coredamage event during shutdown, (1) loss of
function of the residual heat removal (RHR) system, (2) loss of the
external power, (3) loss of function of the reactor coolant pressure
boundary, and
reactivity insertion are assumed, but the "reactivity
insertion" due to a rod withdrawal event etc. accompanying HCU
(hydraulic control unit) isolation work is considered not to be an
initiating event of a "significant core-damage event"
since the effect is limited only to the fuel bundles around the
withdrawn control rod, and even if fuel is failed, the event
locally terminates and does not lead to a large core damage.
Accidental control rod withdrawal during shutdown has been prevented
by operational administrative controls, such as placing all control
rods in full-insertion position during shutdown and using a return
operation to connect the nuclear reactor with the control rod
hydraulic system when many HCUs are to be isolated. But, control rod
withdrawal events were confirmed by some
licensees in 2007, requiring more thorough safety management be put
in place during shutdown.
Japanese BWR electric utilities and manufacturers established a
working group in the Japan BWR Owners Group (JBOG) in June 2007 and
the group has reviewed safety management during shutdown. Following
the occurrence of control rod withdrawal events, the Atomic Energy
Society of Japan has studied whether review of the evaluation
methods is required or not in the periodical revision work for the
"Procedures of Probabilistic Safety Evaluations on Shutdown
State of Nuclear Power Station: 2002."
The safety of nuclear power plants in Japan
is ensured sufficiently by taking strict measures for safety
assurance at each stage of design, construction, and operation. As a
result, the occurrence probability of a severe accident, which is an
event to drastically exceed the design basis events (events that
should be considered in the safety design and evaluation of a
nuclear reactor facility, out of events likely to lead to abnormal
conditions of the facility) and to cause significant damage to the
core, due to conditions in which proper core cooling or reactivity
control cannot be performed with means assumed in the evaluations of
safety design, has become so low that its actual occurrence is
inconceivable.
However, in May 1992, the Nuclear Safety Commission adopted a policy
to further reduce this low risk. In the case of an event beyond the
design basis event that could severely damage the core, in order to
prevent an escalation of the event to a severe accident, or even if
it has escalated to a severe accident, in order to mitigate the
consequences, measures to be taken should be appropriately
performed.
The severe accidents to be assumed for BWRs and the measures to be
taken are provided in the following.
Station blackout event
i) Restoration of external
power or restoration of diesel generators
Reactor scram failure (ATWS) event
i)Manual scram or manual
control rods insertion in the event the reactor protection system is
inoperable
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Psa during Shutdown
Severe Accident
Chapter
7 Safety Evaluation of BWR Plants
Manual startup of the standby liquid control (SLC) system
Loss of decay heat removal function event during transients
Restoration of the residual heat removal system (RHR)
Manual startup of the containment spray system
Containment venting
Failure of water injection after transients
Manual startup of the high pressure ECCS and the reactor core
isolation cooling (RCIC) system
Manual startup of the automatic depressurization system (ADS) and
the low pressure ECCS
Manual startup of alternative water injection facilities
In addition, manual startups of SLC, ADS and ECCS, and actions such
as containment venting and restoration of equipment, etc. out of the
actions described in the above are taken into account for the level
1 PSA. Moreover, for the containment venting, existing facilities
(atmospheric control system or standby gas treatment system) are
considered to be used in the evaluation. The venting in this case is
to prevent containment damage during the event of a loss of decay
heat removal function after transients.
References
“Review Guide for Nuclear Reactor Siting
Evaluation”
“Review Guide for Safety Evaluation of Light Water Nuclear Power
Reactor Facilities”
“Review Guide for Safety Design of light
Water Nuclear Power Reactor Facilities”
"Review Guide for Classification of Importance of Safety
Functions for Light Water Nuclear Power Reactor Facilities”
“Review Guide for Emergency Core Cooling System Performance of
Light Water Nuclear Power Reactors”
“Review Guide for Reactivity Insertion Events of light
Water Nuclear Power Reactor Facilities”
"Meteorological Guide for Safety Analysis of
Nuclear Power Reactor Facilities”
“Guide for Dose Objectives around Light Water
Nuclear Power Reactor Facilities”
"Logic of Nuclear Safety (New Edition)", Sato Kazuo,
Nikkan Kogyo Shimbun, 2006
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