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1 Introduction
The objective of this study is to analyze Steam Generator Tube Rupture (SGTR)
using Event Tree Analysis (ETA). This is a Probabilistic Safety Assessment (PSA)
technique used for identifying potential accident sequences and quantifying risk for
evaluating contribution to Core Damage Frequency (CDF). It includes mitigating actions
interventions in steam generator tube rupture sequences leading to severe accident
conditions. In case of SGTR, current accident management actions foresee flooding of
the secondary side through the emergency feed water system in an attempt to arrest the
activity. Effective accident management actions may significantly reduce the source term
in these accident types. PSA is used to explore the risk significance for various aspects of
plant design or operation and for the evaluation of abnormal events that occur at theplant. It identifies the sequences of events that can lead to core damage, estimates the
core damage frequency and provides insights into the strengths and weaknesses of the
safety systems and procedures provided to prevent core damage. The necessity for these
evaluations is the rationale for establishing a PSA applications program.
Any issue that is going to be evaluated needs to be explicitly defined together with
the type of results required as input to the decision making. As already stated, as part of
the evaluation, the PSA is used in combination with other methods and sources of
information. The PSA can be used to evaluate the risk significance of each issue, or to
define a risk measure as the basis of prioritizing the various issues under consideration
[1]. Having invested considerable resources in developing PSAs, there is a desire to use
the insights derived from them to enhance plant safety while operating the nuclear
stations in the most efficient manner. PSA is an effective tool for this purpose as it assists
plant management to target resources towards the largest risk of accident.
The assessment of risk with respect to nuclear power plants is intended to achieve the
following general objectives:
Identify initiating events and event sequences that might contribute significantly
to risk;
Provide realistic quantitative measures of the likelihood of the risk contributors;
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Provide a realistic evaluation of the potential consequences associated with
hypothetical accident sequences; and
Provide a reasonable risk-based framework for making decisions regarding
nuclear plant design, operation, and siting.
1.1 Role of PSA in NPP Safety Management
During the operation of a nuclear power plant, conditions exist that alter the risk
of operating the facility. These conditions (or events) that result in a change, where
change can be an increase or decrease in risk, fall under three general categories. It
includes plant activities that dictate that certain components will be incapable of
performing their desired functions at certain times during operation [2].
The objectives of a NPP PSA study are as follows:
To determine core damage frequency (CDF) based on the fault trees method and
event trees method,
To identify initiating events and accident sequences with a predominating
contribution to core damage,
To assess the effects of various modifications of safety systems on CDF,
To specify recommendations for the updating of emergency operating procedures
based on predominant accident sequences.
A nuclear power plant PSA analyses the risk associated with operating the plant,
expressed in terms of various postulated initiating events (PIEs) related to the different
levels of damage to the plant (e.g. core damage frequency). The analysis is done using a
logical and systematic approach that makes use of realistic assessments of the
performance of the equipment and plant personnel as a basis for the calculations. This in
principle has the potential to produce an understanding of the inherent risk of operating
the plant over a much wider range of conditions than the traditional deterministic
methods, which generally define what is assumed to be a bounding set of fault conditions.
In international practice three levels of PSA have been evolved:
1. Level 1: The assessment of plant failures leading to the determination of core
damage frequency;
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2. Level 2: The assessment of containment response leading, together with Level 1
results, to the determination of containment release frequencies;
3. Level 3: The assessment of off-site consequences leading, together with the
results of Level 2 analysis to estimate the potential environmental and health
effects.
A Level 1 PSA identifies the sequences of events that can lead to core damage,
estimates the core damage frequency and provides insights into the strengths and
weaknesses of the safety systems and procedures provided to prevent core damage.
Level 2 PSA, which identifies the ways in which radioactive releases from the plant
can occur and estimates their magnitudes and frequency. This analysis provides
additional insights into the relative importance of the accident prevention and
mitigation measures such as the reactor containment. Level 3 PSA, which estimates
public health and other societal risks such as contamination of land or food. This
particular study involves a level 1 PSA.
One of the products of a probabilistic risk assessment (PRA) is a list of plant
responses to initiating events (accident starters) and the sequences of events that
could follow. By evaluating the significance of the identified risk contributors, it is
possible to identify the high-risk accident sequences and take actions to mitigate
them. Although the consequences of the high-risk accident sequences may vary from
one PRA to another, they all attempt to evaluate realistically the consequences of
hypothetical accident sequences.
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2 Init iating Event Analysis
2.1 Initiating Events
An initiating event (IE) is a postulated event that could occur in a nuclear power
plant. It is an occurrence that creates a disturbance in a plant and has the potential to lead
to core damage, depending on the successful operation or failure of the various mitigating
systems in the plant. An initiating event is an incident that requires an automatic or
operator initiated action to bring the plant into a safe and steady-state condition, whereas
in the absence of such action the core damage states of concern can result in severe core
damage.
Initiating events are usually categorized in divisions of the internal and external initiatorsreflecting the origin of the events.Initiating events are generally classified into internal
IEs and hazards (internal and external). Internal IEs are hardware failures in the plant or
faulty operations of plant hardware through human error or computer software
deficiencies. External hazards (which may also be termed external events) are events that
create extreme environments common to several plant systems. External hazards include
earth- quakes, floods, high winds and aircraft crashes. Internal hazards include internal
flooding, fire and missile impact [3].
2.1.1 Type Of Initiating Events
The internal initiating events are categorized as follow [4]:
1) LOCAs Initiators; The loss of coolant accident (LOCA) initiators include
primary system breaks resulting in loss of primary coolant. Pipe breaks and
ruptures of different sizes, inadvertent opening and failures to re-close (stuck
open) of valves are being considered in this category.
2) Transients Initiators; The transient initiating events are those which
introduce the disturbance in normal plant operation, without loss of primary
coolant and which require an automatic or manual shutdown of the reactor.
Examples of transients are: disturbance in feed water flow of turbine/condenser,
reactivity control, reactor re-circulation, etc.
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3) Common Cause Initiators (CCIs); These are events, which, in addition to
requiring reactor shutdown, simultaneously disable one, or more of the mitigating
systems required to control the plant status following the initiator.
The important types of LOCA Ies and CCIs are represented in tables 1and 2 respectively.
Table 1: List of important LOCA initiators [4].
Events Description of Breaks
1. Small-break LOCA A break or leak 1/2 to 4 inches in effective diameter.
These are spontaneous events: induced LOCAs were
treated directly.
2. Large LOCA A break or rupture greater than 4 inches in effectivediameter except those noted below.
3. Interfacing-system
LOCA
A large loss of coolant through the valves acting as a
Boundary between high and low RCS pressure
4. RPV rupture A loss of reactor-vessel integrity precluding the ability toMaintain coolant inventory.
5. Steam generatortube rupture
A rupture of a steam generator tube resulting in an RCSleak greater than 10 gpm.
Table 2: List of important Common Cause initiators [4].
Events Description of Breaks
1) Loss of instrument-air system
Reactor trip, a failure of instrumentation and equipmentdue to this system.
2) Loss of service-
water system
A pipe break or pump failure and in addition prevent
other safety system operation that depend on service
water cooling supply.
3) Loss of integrated
and auxiliary controlsystems
The Integrated Central System (ICS) is controlling feed
water, pressurizer heaters etc., and may cause a transientwith loss of a protecting system.
4) Loss of DC power
system
Failure to supply power to a number of pumps in train A
Supplying several mitigating systems.
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2.2 Selection Of Initiating Events
A long list of initiating events (IEs) is recognized completely as possible. A judgment is
required that any IEs not identified would make only a small contribution to the total risk.
The scope of the PSA could also constrain the initiating events that are to be considered.
There are several approaches to the selection of IEs, each of which has its limitations.
Since the aim is to produce a list that is as complete as possible. . The approaches are
discussed below [3].
2.2.1 Engineering Evaluation
This technique is directly related to evaluation of a component of plant system.The plantsystems (operational and safety) and major components are systematically reviewed to
evaluate the failure modes, for example (failure to operate, spurious operation, breach,disruption, collapse) that could lead directly, or in combination with other failures, to
core damage. Partial failures of systems should also be considered since, although they
are generally less severe than complete failure, they are of higher frequency and are often
less readily detected.
2.2.2 Reference To Previous Lists
This technique refers to studies of lists of IEs drawn up for previous PSAs on similar
plants and for the safety analysis report. This may in fact be the starting point. Specially
useful for providing a list of transient initiators for LWRs.
2.2.3 Deductive Analysis
This approach uses master logic diagram, in which core damage is made the top event in
a diagram, which has the appearance of a fault tree (although it is not one in the usual
sense). This top event is successively broken down into all possible categories of events
that could cause it to occur. Successful operation of safety systems and other preventive
actions are not included. The events at the most fundamental level are then candidates for
the list of IEs for the plant.
2.2.4 Operational Experience
The operational history (if any) of the plant in question and of similar plants elsewhere is
reviewed for any events that should be added to the list of IEs. This approach is
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supplementary and would not be expected to reveal low frequency events, but it could
show common cause IEs.
2.3 Initiating Events GroupingSince some of the initiating events would induce the same or a reasonably similar plant
response. In order to get rid of a long list of initiating events, these are divided into
groups of IEs. Initiating events can be grouped in such a way that all events in the same
group impose essentially the same success criteria on the front line systems as well as the
same special conditions (challenges to the operator, to automatic plant responses, etc.)
and thus can be modeled using the same event/fault tree analysis [50P-4]. Example is:
steam line break by size, loss of flow by number of pumps failed and spurious control rod
withdrawal by number of rods or rate of reactivity addition events are grouped into one
group.
LOCA IEs are divided into groups on the bases of pipe breakage, leakage or rupture. The
LOCA groups are:
I. Large LOCA; For break size > 6 inch diameter, equivalent to > 300 cm2 leakage
area);
II. Medium LOCA; For break size > 3 inch diameter, equivalent to 150-300 cm2
leakage area);
III. Small LOCA (e.g. break size < 3, equivalent to < 150 cm2 leak area).
Transient Initiators are plant specific, and depend heavily on the purpose and scope of a
PSA study. Transient IEs are divided into groups based on design and operation features
of plant and PSA requirement. The minimum groups are:
I. Transients with main feed water (MFW) initially available (turbine/reactor trips);
II. Transients with loss of MFW;
III. Loss off-site power
Common cause IEs are very much plant specific. The most common initiating groups are:
I. Loss of vital AC power bus;
II. Loss of service water system;
III. Loss of component cooling;
IV. Loss of a DC bus;
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V. Loss of instrument air;
VI. Loss of core level measuring instrument
VII. Loss of ventilation system;
VIII. Loss of room coolers;
IX. Steam line break in locations where it causes additional effects or containment
isolation
2.4 Initiating Events In PWRsOne of the products of a nuclear power plant PRA is a list of plant responses to initiating
events (accident starters) and the sequences of events that could follow. By evaluating the
significance of the identified risk contributors, it is possible to identify the high-risk
accident sequences and take actions to mitigate them.
The barriers confining the radioactivity are manifold. The first is the ceramic fuel pellet
itself; radioactivity must diffuse from the pellet. It is confined within the primary cooling
loop and if released through, for example, the safety relief valve action, it is confined by
the containment. Most probable initiating events in PWRs are depicted in table 1. Nuclear
power plant systems may be classified as "Frontline" and "Support" according to their
service in an accident.
Frontline systems are the engineered safety systems that deal directly with an
accident; and
Support systems provide the services necessary for the frontline system to
function.
List of typical font line systems is given in table 2.Accident initiators are broadly grouped
as loss of cooling accidents (LOCA) or transients. A LOCA is one in which the water
cooling the reactor is lost due to irreversible damage to the boundary holding the water.
These are typically classified as small-small, small, medium and large [5]:
1. SSLOCAs(small-small LOCA), ranging in pipe break sizes up to 3 inches, are
mitigated by high pressure injection from typically one of three pumps,
2. SLOCA(Small LOCA), encompassing pipe break size in the range of 1 to 8
inches are mitigated by high pressure injector from two out of three pumps
and two out of three accumulators.
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3. MLOCAs (Medium LOCA) in the range 6 to 18 inches are mitigated by two
out of three accumulators and one out of two low pressure pumps,
4. LLOCA (Large LOCA) encompassing the largest pipes in the plant is
mitigated by the accumulators and one out of three low pressure-high volume
pumps.
A transient, as the name signifies, is a passing event, which may upset the reactor
operation, but itself does notcause immediate damage
Table 3: List of PWR Transient Initiating Events [5]._________________________________________
1. High pressurizer pressure2. Inadvertent safety injection signal
3. Containment pressure problems
4. Startup of inactive coolant pump
5. Total loss of RCS flow6. Loss or reduction in feed water flow (one loop)
7. Total loss of feed water flow (all loops)
8. Full or partial closure of MSIV (one loop)
9. Feed water flow instability-miscellaneous mechanical causes
10. Loss of condensate pumps (one loop)
11. Loss of condensate pumps (all loops)
12. Steam-generator leakage
13. Sudden opening of steam relief valves
14. Turbine trip, throttle valve closure
15. Generator trip or generator-caused faults
16. Loss of all offsite power
17. Pressurizer spray failure
18. Spurious trips-cause unknown
19. Manual trip-no transient condition
20. Fire within plant________________________________________________________________________
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3 Event Tree Analysis
Event trees are graphic models that order and reflect events according to the requirements
for mitigation of each group of initiating events. Events or headers of an event tree can be
a safety function's status, a system's status, basic events occurring or operator actions.
Event trees display some of the functional dependences between the events or 'headings'
of the tree; e.g. cases where failure of one system implies that another system cannot
perform its function successfully. Such dependences result in omitted branch points.
Omitted branch points also occur if the failure of a given system does not affect the plant
damage state associated with a given accident sequence.
The event tree headers are normally arranged in either chronological or causal order.
Chronological ordering means that events are considered in the chronological order inwhich they are expected to occur in an accident. Causal ordering means that events are
arranged in the tree so that the number of omitted branch points is maximized [3].
The event-tree method is described as a method for modeling plant-level sequences that
may lead to public risk. The approach to event-tree development and application is
generalized and can be adapted to specific study objectives. The event-tree method has
been used in some form in all recent risk assessments for light-water reactors. It is a most
suitable means for modeling complex plant-level sequences, and it permits these
sequences to be evaluated in an efficient manner.
The integration of event trees and fault trees provides an analytical approach capable of
handling the complexities associated with modeling potential accident sequences. It is a
proved means for defining and under- standing plant design and operation in a manner
that leads to the quantification of public risk [6].
Quantification of the risk associated with a commercial nuclear power plant requires the
delineation of a large number of possible accident sequences. Because nuclear systems
are complex, it is not feasible to write down a listing of important sequences. A
systematic and orderly approach is required to properly understand and accommodate
many factors that could affect the course of potential accident.
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Figure 1: Procedure for the even tree development [6].
3.1 Event Sequence Analysis
Event sequence analysis is a method used to identify the complex relation ships between
accident-initiating events and detailed system responses. Event sequence diagrams
(ESDs) are developed for each group of the initiating events. The ESD is an analytical
tool intended to facilitate the collection and display of information required for the
developing system of event trees. Its objective is to illustrate all possible success paths
from a particular accident-initiating event to a safe shutdown condition.
3.1.1 Success Criteria
It is the criteria that have been developed for mitigating the events that constitute core
damage. This is often done by adopting indirect criteria where core damage is assumed to
occur following prolonged core uncovery, to the top of the core or over pressurization
and these need to be differentiated for comprehensive analysis. This is often assumed for
light water reactors but is not necessarily applicable for all reactor types. The safety
functions that need to be performed to prevent core damage are to be identified for each
of the initiating event groups. The safety functions required would typically include
detection of the initiating event, reactor shutdown, residual heat removal, containment
protection, etc. depending on the nature of the initiating event. The safety systems
available to perform each of these safety functions have to be identified [10].
Definitionof
Safety
functions
SelectionOf
Initiating
events
Evaluationof
Plant
response
PlantFamiliarizat-
-ion
DelineationOf
Accident
sequences
System
Modeling
tasks
Evaluationof
plantdamage
state
Identificationof
system
failure
criteria
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The success criterion for each system is then determined, as the minimum level of
performance required from the system, and expressed.
3.2 System Modeling
A general objective of risk assessment is to determine the susceptibility of a system or of
groups of systems to condition of design, operation, test, and maintenance that could lead
to failure. This objective can be realized through system modeling, for which a variety of
analytical techniques can be used.
The level of the PSA determines some of the factors that must be accounted for in the
system models. Information on the elevation of a component, proximity to specific
systems or components, or room location with in the plant is typical of the information
needed for system model ling if floods, fires, earthquakes, or similar external hazards are
to be properly addressed. Decisions also are required as to the level pf detail and the type
of components to be included in the trees. Normally, passive failure of piping segments
are omitted or lumped together. If the segments and information on their location are
included. Figure 2 shows the generalized process of system fault tree modeling. A
significant amount of system related information is generated during the plan-
familiarization process. This information, along with specific system failure criteria
developed for each of event tree heading forms the basis for the system modeling.
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Figure 2: Generalized process of system modeling [6].
The initial step is the definition of the top events for each fault tree, these must be
consistent the appropriate event tree heading. When the top event has been clearly
defined, the ground rules for analysis must be clearly specified. The system under
analysis must be clearly defined and its boundaries and interfaces identified. The
constraints and assumptions associated with analysis must be under stood and
incorporated into the model 6].
AccidentSequence
Quantificatio
n
Developmentof
System
fault trees
Specificationof
Analysis
Ground
rules
DefinitionOf
Fault tree
Top
Events
Identificationof
System
Failure
Criteria
Preparationof
Fault trees
For
evaluation
Developmentand
Application
Of numerical
data
PlantFamiliar-
-zation
3.3 Safety Functions
The functions that must be performed to control the sources of energy the plant and the
radiation hazard are called safety functions. The concept of safety functions forms the
basis for selecting accident initiating events and delineating potential plant responses.
Generally, safety functions are defined by a group of actions that prevent core melting,
prevent containment failure ,or minimize radio nuclide release. Such actions can result
from the automatic or manual actuation of a system, from passive system performance, or
from the natural feedback inherent in the design of the plant [6].
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For each IE, the safety functions that need to be performed in order to core damage is
identified [3]. Nuclear power plant systems may be classified as "Frontline" and
"Support" according to their service in an accident. Frontline systems are the engineered
safety systems that deal directly with an accident while Support systems provide the
services necessary for the frontline system to function [5]. The important safety functions
are listed in table 3.
3.3.1 Frontline systems
Frontline systems are the engineered safety systems that deal directly with an accident in
the plant. Examples of front line systems for a PWR are:
Reactor protection system
Core flood system
High pressure injection/re-circulation system
Low pressure injection/re-circulation system
Reactor building spray injection/re-circulation system
Reactor building cooling system
Power conversion system
Emergency feed water system
Pressurizer safety relief valves
3.3.2 Support systems
The systems that are required for the proper functioning of the front line systems are
termed support systems. Their performance as a safety function is indirect.
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Table 4: Safety function and front line system corresponding to a particular initiating event.
Initiating event Safety function Frontline systems
LOCA Render reactor sub-
critical
Remove core decay heat
Prevent containment over
pressurization
Scrub radioactive
materials
Reactor protection system
High pressure injection systemLow pressure injection systemHigh pressure re-circulation system
Core flood tanks
Auxiliary feed water system
Power conversion system
Reactor building spray injection system
Reactor building spray re-circulationsystem
Reactor building spray fan cooling
systemIce condensers
Reactor building spray injection systemReactor building spray re-circulation
systemIce condensers
Transients Render reactor subcritical
Remove core decay heat
Prevent containment overpressure
Scrub radioactive
materials
Reactor protection system
Chemical volume and control
High pressure injection System
Auxiliary feed water system
Power conversion system
High pressure injection systemPower-operated relief valve
Containment spray injection systemContainment spray re-circulation
system
Containment spray fan cooling system
Ice condensers
Containment spray injection system
Containment spray re-circulation
systemIce condensers
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4 Steam Generator Tube Rupture Event
An SGTR event is a loss-of-coolant accident that results in a leakage of the primary
coolant into the secondary side of one or more (SGs). This type of event poses several
rather unique operational concerns such as: steaming of a ruptured SG results in offsite
radiological doses, a continuous in-leakage results in SG overfill, and failure to reduce
the differential pressure between the primary and secondary sides can result in the
depletion of the borated water storage tank (BWST) inventory. A leakage rate of primary
coolant would depend on the severity of tube rupture and may vary from several gallons
per minute (gpm) in the case of a single tube failure, to several hundreds to thousands of
gpm in the case of guillotine rupture of several tubes.
The accident is assumed to take place at power with the reactor coolant contaminated
with fission products corresponding to continuous operation with a limited amount of
defective fuel rods. The accident leads to an increase in contamination of the secondary
system due to leakage of radioactive coolant from the SRC. In the event of a coincident
loss of offsite power, or failure of the condenser steam dump system, discharge of
activity to the atmosphere takes place via the steam generator safety and/or poweroperated relief valves.
Complete severance of a steam generator tube is considered a some what conservative
assumption since the Incoloy 800 tube material is highly ductile. The more probable
mode of tube failure would be one or more minor leaks of undetermined origin. Activity
in the steamand power conservation system is subject to continuous surveillance and an
accumulation of minor leaks which exceed the limits established in the technical
specification is not permitted during the unit operation [8].
In case of SGTR, plant conditions are defined in terms of general accident scenario and a
five critical safety functions: primary pressure control, primary inventory control,
secondary heat sink, secondary pressure control and secondary heat removal [11].
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4.1 Purpose of analysis of SGTRThe analyses of steam generator tube rupture (SGTR) event are performed to evaluate the
following scenarios [7]:
An SGTR transient with leak rate less than normal makeup rate (less than a single
tube rupture) and leak rate greater than normal makeup rate capacity,
Steaming of both SG versus isolation of an affected SG,
Breaks in both SGs,
Off-site power available and loss of off-site power,
Steam Generator Tube Rupture (SGTR) is an initiating event considered in PWRs
only. In this project only one tube rupture in a steam generator is considered. Even
though this is a very small loss of coolant accident (LOCA) the plant response is in
general different from the very small LOCA case (due to filling of affected SG and
eventually over pressurizing it) and, in addition, a path to bypass containment is created
in this case, which makes this initiator unique [4].
To estimate the core damage frequency, a small event tree and large fault tree PRA
technique is used. The event trees are used to simulate the procedure, while the fault treesare used to simulate the systems called out in the event trees to prevent the core damage.
The sequences are developed and quantified using the Integrated Reliability and Risk
Analysis. Every effort is taken to eliminate conservative PRA modeling assumptions.
For example, a "failure to depressurize" event is not assumed to result in fuel damage,
given that high pressure injection (HPI) pump is available. Similarly, all the efforts are
taken to preserve simplicity and understanding of the models by eliminating unwarranted
complexity. The event trees are very large, complex, and consist of large number of
sequences, e.g., 137 sequences for a less than single tube rupture event compared to 10 to
15 sequences in a conventional SGTR event tree [9].
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4.2 Future Tasks
The future work would be totally dedicated to the detailed analysis of steam generator
tube rupture (SGTR) event and to evaluate the operation of safety functions of Nuclear
Power Plant system to mitigate this event. Safety function are analyzed under the headingof even tree header e.g. high pressure injection system, low pressure injection system,
and residual heat removal system.
The scope of the project that comprises the major part of the project, is quantification of
the initiating event of steam generator tube rupture in a PWR core. The planning of the
project for fifth semester would consist of following points:
Use Of Risk Spectrum Professional
To develop SGTR event tree
To create respective fault trees
Linking fault tree top gates to event tree headers
Accident Sequence Analysis
Accident Sequence Quantification
Interpretation of Results
Identification of most severe accident sequences and top minimal cut sets
Contribution of SGTR in total CDF
Discussion of results
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5 Summary and Conclusions
The objective of this study is to analyze Steam Generator Tube Rupture (SGTR)
initiating event using Event Tree Analysis (ETA). Probabilistic Safety Assessment (PSA)
technique is used for identifying potential accident sequences and quantifying risk for
evaluating contribution to Core Damage Frequency (CDF). A Level 1 PSA identifies the
sequences of events that can lead to core damage, estimates the core damage frequency
and provides insights into the strengths and weaknesses of the safety systems and
procedures provided to prevent core damage. Level 2 PSA identifies the ways, in which
radioactive releases from the plant while Level 3 PSA estimates public health and other
societal risks such as contamination of land or food.
An initiating event (IE) is a postulated event that could occur in a nuclear power plant. Itis an occurrence that creates a disturbance in a plant and has the potential to lead to core
damage. Initiating events are categorized into LOCAs, transients and common cause
failures. The several approaches for the selection of initiating events are engineering
evaluation, deductive analysis, operational experiences and reference to previous lists.
The initiating events are divided into groups to get rid of large number of events in a
Nuclear Power Plant on the bases of same initiating conditions. Initiating events are
grouped in such a way that all events in the same group impose essentially the same
success criteria on the front line systems as well as the same special conditions. The
integration of event trees and fault trees provides an analytical approach capable of
handling the complexities associated with modeling potential accident sequences. It is a
proved means for defining and under- standing plant design and operation in a manner
that leads to the quantification of public risk. Quantification of the risk associated with a
commercial nuclear power plant requires the delineation of a large number of possible
accident sequences. A general objective of risk assessment is to determine the
susceptibility of a system or of groups of systems to condition of design, operation, test,
and maintenance that could lead to failure. This objective can be realized through system
modeling, for which a variety of analytical techniques can be used. Safety functions are
defined by a group of actions that prevent core melting, prevent containment failure, or
minimize radio nuclide release, which are front line systems and support systems to
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mitigate the particular event happened in the reactor. An SGTR event is a loss-of-coolant
accident that results in a leakage of the primary coolant into the secondary side of one or
more steam generators. The accident is assumed to take place at power with the reactor
coolant contaminated with fission products corresponding to continuous operation with a
limited amount of defective fuel rods.
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References
[1] International Atomic Energy Agency, Applications of Probabilistic Safety
Assessment (PSA) for Nuclear Power Plants IAEA-TECDOC-1200,
International Atomic Energy Agency, Vienna, 2001.
URL:http://www.energyrisks.jrc.nl/APSA/PDF/Publications/Useful%20reference
s/IAEA%20TECDOC%201200.pdf
[2] Smith,C, Borgonovo,E, George Apostolakis, Review of International Activities
in Accident Management and Decision Making in the Nuclear Industry, May,
1999,Massachusetts Institute of Technology.
[3] International Atomic Energy Agency, Procedures for conducting probabilistic
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