4th Semestr Report

8/12/2019 4th Semestr Report

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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

earthquakes, 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 understanding 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


High pressure injection systemLow pressure injection systemHigh pressure re-circulation system

Core flood tanks

Auxiliary feed water 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


Transients Render reactor subcritical

Remove core decay heat

Prevent containment overpressure

Scrub radioactive

materials


Chemical volume and control

High pressure injection System

Auxiliary feed water 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



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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 understanding 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

safety assessment of nuclear power plants (level-1), safety series 50p-4,

International Atomic Energy Agency, Vienna, 1992.

[4] International Atomic Energy Agency, Defining initiating events for purposes of

Probabilistic safety assessment, TECDOC-719, A-1400, September1993,

Vienna, Austria.

[5] Hall.E.R, Fullwood.R.R, Probabilistic Risk Assessment In The Nuclear Power

Industry: Fundamentals And Application, 3rdedition, 1998, Brook Heaven

National Laboratory, New York, USA.

[6] PRA Procedures Guide, A Guide to Performance of Probabilistic Risk

Assessment Of Nuclear Power Plants NUREG/CR-2300, vol.1, New York,1983.

[7] International Atomic Energy Agency, Review of Probabilistic Safety

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