PROBABILISTIC SAFETY ASSESSMENT OF

NARORA ATOMIC POWER PROJECTby

A. K. Babar, R. K. Saraf, A. Kakodkar Reactor Engineering Division

andV. V. S. Sanyasi Rao

Health Physics Division

1989

B.A

.R.C

. - 14

45

B.A.R.C. - 1445

GOVERNMENT OF INDIA ATOMIC ENERGY COMMISSION

PROBABILISTIC SAFETY ASSESSMENT OF NARORA ATOMIC POWER PROJECT

byA.K. Babar, R.K. Saraf and A. Kakodkar

Reactor Engineering Divisionand

V.V.S. Sanyasi Rao Health Physics Division

BHABHA ATOMIC RESEARCH CENTRE BOMBAY, INDIA

1989

INIS Subject Category s E34.00

Descriptors

NARORA-1 REACTOR NARORA-2 REACTOR REACTOR SAFETY PROBABILISTIC ESTIMATION FAILURE MODE ANALYSIS REACTOR ACCIDENTS MONTE CARLO METHOD FAULT TREE ANALYSIS RELIABILITY FAILURESREACTOR COOLING SYSTEMSPUMPSFEED WATERECCMODERATORSVALVESPIPESCALANDRIAS CONDENSATES HEAT EXCHANGERS COMPRESSORS POWER SUPPLIES REACTOR SHUTDOWN LEAKSCONTAINMENT

EBDBABILISIIC-SAEEJDL-ASSESSHEHIOE

NARDRA ATOMIC POWER PROJPCT

ABSTRACT

Various safety studies on Pressurised Water. and Boiling Water reactors have been conducted. However, a detailed report on probabilistic safety assessment(PSA) of PHWRs is not available. PSA level T results of thestandardised 235 MWe PHWR under construction at Narora are presented herein. Fault Tree analysis of various initiating events(IEs), safety systems haa been completed. Event Tree analysis has been performed for. all the dominating IEs to identify the accident sequences and a list of thedominating accident sequences is included. Analysis has been carried out using Monte Carlo simulation to propagate the uncertainties in failure rate data. Further, uncertainty analysis ic extended to obtain distributions for the accident sequences and core damage frequency. Some noteworthy results of the study apart from the various design modifications incorporated during the design phase are : .

i) The accident sequences resulting from station blackout are dominant contributors to the core damage frequency.

ii) Class-IV transients, small break LOCA are significantIEs. Main steam line break is likely to induce (staeam S-Uc"4' generator tube ruptures.

iii) Moderator circulation, fire fighting system, secondary steam relief are relatively important in core damage frequency reductions.

iv) Under accidental situations human errors are likely to be associated with valving in shutdown cooling and fire fighting systems.

(i>

LEOEHDIE -Initiating Event

ESF -Engineered Safety Function

ML -Medium LOCA

SL -Small LOCA

Class JV -Main( Grid & Station) Power Supply

Class III -Emergency(DG) Power Supply

FW -Feed Water

MSLB -Main Steam Line Break

APWS -Active Process Water System

NAHPPWS -Non-active High Pressure Process Water System

ECI -Emergency Coolant Injection

ECR -Emergency Coolant Recirculation

RBI -Reactor Building Isolation

RBC -Reactor Building Cooling

RT -Reactor Trip

SLHS -Small LOCA Handling System

FFS -Fire Fighting System

SSR -Secondary Steam Relief

AFWS -Auxiliary Feed Water System

SDC -Shutdown Cooling System

(ii)

EAPWS -Emergency Active Process Water System

ЯЕ -Human Error

CCF -Common Cause Failure

PROBABILISTIC SAPETT ASSESSMEt QZ

HARQBA ATOMIC POWER project

<¥4Л»PPCIIOH

Probabiliatic Safety AsaeasaentCPSA) in the context of Nuclear Power Plants(NPPs) is associated with the nodels that predict the offsite radiological release resuiting from the potential reactor accidents. In its entirety, it comprises the following levels:

1.Identification of accident sequences and quantitative estimates of the frequency of each i.e System Analysis

2. Radiological release to the environment associated with each class of accident sequence i.e. Containment Analysis

3. Analysis of the off-site consequences of the releasei.e. Consequence Analysis

It is intended to obtain a fullscope probabilistic model of a standardised 235MWe Pressurised Heavy Water Reactor (PHWR) which would be uséd in the safety and operational analysis of the reactor. The model would be a risk management tool to meet the following objectives.

a. Determining the core damage frequency using a set of internal Initiating Events(IEs) and external IEs like loss of off-site.power

b. Identification and quantification of the dominating! accident sequences, uncertainities and spécifié contributers to system failures to establish their teat: and procurement procedures

c. Identifying design and operational weaknesses

d. Supporting decisions on safety issues

e. Developing test and maintenance schedules anddetersuning allowable outage times to assist in the establishment of criteria for Technical Specifications

f. Correlating accident »equences to release categories

g. Consequence modelling and risk estimation

Narora Atomic Power Project(KAPP) is a 235 KVe PHVR under construction and would be a standardised design for the forthcoming similar projects under construction or being planned. A PSA study of KAPP was undertakeri to identify the dominating accident sequences relevant to PHWR design, quantify the same using System Reliability Models like Fault Tree etc. to fulfill the various objectives and perform a design evaluation to improve Safety and Reliability and possibly think in terms of an Inherently Safe Reactor. This report presents the results of Level-I PSA carried out for HAPP in terms of the following information.

a. Identification of dominating Initiating Events

b. Reliability analysis of various IEs and the Engineered Safety Functions(ESFs) using Fault Tree methods.

c. Identification of accident sequences leading to core damage using Event Tree methods

d. Quantification of accident sequences to obtain dominating accident sequences leading to fixing the reliability requirements of various systems in reducing the risk

e. Uncertainty analysis and error propagation to account for the variability in component failure data, accident sequence and core damage frequency etc.

3LimiIAIXHC ITS

Many iapoctant studies.exaaplea [13,t2] have been perforeed on the use of PSA In case of Light Water Reactors(LWRs), however, a detailed study is yet to appear for a PHVR. In order to Identify the IEs applicable to a PHWR, it would be worthwhile to list the different design features. The PHWR is a heavy water cooled, heavy water ■oderated, natural uraniun fuelled reactor which utilises the pressure tube concept. The pressure tubes containing the fuel run horizontally through the reactor core. Each pressure tube is isolated and insulated froa the heavy water aoderator by a concentric calandria tube and a gas annulus. The aoderator is operated at low teaperature and pressure. The reactivity control and shutdown aechanisas reside in the low pressure aoderator, thus siaplifying their design, construction and áaintenance and eliainating virtually, the possibility of their ejection in an accident situation. In the standardised design, two fast acting, independent, diverse shutdown systeas are provided and on a reactor trip, the aoderator is not duaped. Thus, in case of loss of coolant accidents, the cool.aoderator can act as a heat sink.

The IEs can be generally classified into the following aain groups:

1. Decrease in reactor coolant inventory

2. Increase in reactor coolant inventory

3. Decrease in reactor coolant system flow rate

4. Decrease in heat renoval by secondary systea

5. Increase in heat renoval by secondary systea

6. Reactivity and power distribution anoaalies

7» Anticipated transients without scranUTWS)

8. Radioactive releasee from a sub-system or component

9. Others

Annex. 2 of Safety Guide SGD11[3] gives a list of IEs generally analysed for the application of a licence for LVR in U.S.A. A number of IEs listed below were added to account for the design differences between PHWRs and LWRa.

1. Leakage from the seal plug after refuelling (group 1)

2. Bleed valve stuck open(l)

3. Failure of a limited number of tubes in any heat exchanger other than steam generator in PBT systea(1)

4. Failure of coolant channel including its end fitting(l)

5. Feed valve stuck open(2)

6. Bleed valve stuck closed(2)

7. Bleed isolation valve closed(2)

8. Flow blockage in any coolant channel assembly/any feeder(3)

9. Failure of reactor moderator flow(6)

10. Failure at any location of moderator system piping(6)

11. Failure of fuelling machine when off the reactor and full of irradiated fuel(8)

A composite list incorporating the IEs given in reference [3] and those enumerated above was prepared and is given below.

1.0 Increase in heat removal by .Bccondarv_gygl:em

1.1 Feed water system malfunction that results in decrease in feed water temperature.

1.2 Feed water system naifunction that resulta in an increase in feed water flow.

1.3 Steam Pressure Regulator(Regulating eyeten)nalfunction or failure that results in increasing stean flow.

1.4 Inadvertant opening of a relief valve resulting in stean flow increase.

1.5 Spectra of steam system piping failures inside and outside containment.

2.0 Decrease in heat removal bv the secondary system2.1 Boiler pressure control(BPC) system nalfunction

resulting in decrease in stean flow.

2.2 Loss of external electrical load.

2.3 Turbine trips

2.4 Inadvertant closure of main steam isolation valve.

2.5 Loss of condenser vaccun.

2.6 Class IV power failure i.e. coincident loss of station as well as grid supply.

2.7 Loss of nornal feed flow.

2.0 Feed water piping break3.0 Decieftge .in reactor coolant .aydten flow ifttu

3.1 Single and multiple reactor coolant pump trips.

3.2 Coolant punp shaft seizure.

3.3 Coolant punp shaft breakage.

3.4 Flow blockage in any reactor fuel channel assembly.

3.5 Failure of all mechanical seals on PHT pump(s).

4.0 Increase in reactor coolant inventory.

4.1 Feed valve stuck open.

4.2 Bleed valve stuck closed.

4.3 Bleed isolation valve closed by mistake by the operator.

3.0 Decrease in reactor спо1аП^ inventory.

5.1 Inadvertant opening of a relief valve in PHT system.

5.2 Feed water tube or insturment tube breakage.

5.3 Steam generator tube/tubes failure.

5.4 End plug fails to close after refuelling.

5.5 PBT header and piping failure.

5.6 Bleed valves stuck open.

5.7 Feed isolating valve closed by operator's mistake.

5.8 Pressure tube failure( followed by calandria tube failure releasing PHT coolant to the moderator.

5.9 Failure of large number of tubes in any heat exchanger( other than steam generator) in PHT system( bleed cooler, gland cooler, shutdown cooler).

5.10 Failure of end fitting of any channel assembly followed by the failure of lattice tube of end shield through which the end fitting runs.

5.11 Failure of mechanical joint between pump cover and pump casing of main coolant pumps.

5.12 Massive failure of a pump cover/casing of main coolant pump.

6.0 Eeactivitv_and power distribution anamollaa.

6.1 Uncontrolled withdrawal of control rod( Reactivitycontrol nechanien) assembly from a sub-critical or low power start up condition( assuming the mostunfavourable conditions of the core and reactor coolant system).

6.2 Uncontrolled withdrawal of control rod assembly at a particular power( assuming the most unfavourable reactivity conditions of the core and the reactor coolant system) that yields the most severe result( low power to full power).

6.3 Chemical control( composition) system malfunction that results in a decrease in boron concentration in reactor coolant.

6.4 Fuel bundle ejection accident.

6.5 Failure of reactor moderator flow.

6.6 Failure at any location of any pipe of reactor moderator system.

6.7 Drop of a load on reactivity mechanisms.

7.0 EftflioacUve release, f roa, ft .cubs va tea pi component.7.1 Tritium leakage.

7.2 Radioactive gas waste system leak or failure.

7.3 Radioactive liquid waste system leak or failure.

7.4 Postulated radioactive releases due to liquid tank failures.

7.5 Design basis fuel handling accident.

7.6 Accidental dropping of spent fusl casks( during transfer of fuel to reprocessing plants).

7.7 Failure of fuelling machine when off-reactorcontaining full complement of irradiated fuel.

7.8 Failure of containment dousing(a) A douse has occured prior to accident.(b) Dousing system is unavailable following accident.

7.9 Containment and associated system failure.

7.10 One door open of air lock or transfer chamber most critical for radioactive release from containment and seals on second door deflated( its impact, for example, when PHT system is leaking or has broken).

7.11 Failure to close any containment isolation device.

8.0 Anticipated transients.without scram( Dual failures).

8.1

8.2

8.3

8.4

8.5

8.6

8.7

Inadvertant withdrawl of control rod( like 6.1 and 6.2 plus faulure of trips).

Loss of feed water.

Loss of class IV power.

Loss of electrical load.

Loss of condenser vaccum.

Turbine trip.

Closure of main steam line isolation valve.

9.0 Others•

9.1 Failure of instrument air.

9.2 Design basis fire.

9.3 Design basis earthquake.

9.4 Degraded operation of containment atmosphere cooling equipment( coupled with PHT failure).

9.5 Leaking containment( coupled with radioactive release from any other systems).

9.6 Turbine overspeed protection faulure.

9.7 Turbine break up.

9.8 Design basis tornado.

9.9 Failure of steam generator support.

9.10 Massive failure of station cooling water tunnel/discharge duct.

Based on the analytical study of the causes and consequences, the following events are considered important for further studies.

1. PHT header and piping failure(group 1)

2. Steam generator tube(s) failure(l)

3. Coolant channel failure(s)(1)

4. Spectrum of steam system piping failure inside and outside containment(5)

5. Loss of normal feed flow(4)

6. Feed water pipe breaks(4)

7. Class XV failure i.e. coincident loss of station as well as grid supply(4)

8. Compressed air failure

9. Fuelling machine induced LOCAs(l)

10. Leakage from the seal plug after refuèlling(1)

11. Loss of regulation

12. Flow blockagè in any coolant channel aaaenbly/feederO)

13. Process water system failure(9)

14. Single and multiple reactor coolant pump failure(s)(3)

15. Failure of a limited number of tubes in any heat exchanger other than the steam generator in PUT system(1)

16. Failure of moderator flow(6)

17. Turbine trips(4)

As can be inferred from the list above, the effect of internally generated missiles, man induced events( air craft crashes) and natural phenomena on the reactor and its associated systems is not considered in this analysis. In addition some events like ATWS are not considered due. to two independent fast acting and diverse shutdown systems in PHWRs. Turbine trip is covered by other events( partly by class IV failure and partly by IRV opening and/or secondary steam relief). Failure of moderator flow is not important as an initiating event. However moderator system is important in those situations where core cooling ' is impaired due to failure of other means of cooling. The remaining events are analysed regarding their frequency and possible impact on the core depending upon the operability states of the various ESFs provided, in the subsequent sections.

3..RSLIABHiII3LAHAbySISIt is important to differentiate between different

categories of systems from the reliability viewpoint. IEs are associated with failure in Process Systems which are active during normal functioning of the reactor e.g. Reactor Regulating System, Primary Heat Transport, fuel etc. where as ESFs are Protective and Containment Systems which are not active during the normal reactor operation but act following failure of a process system to limit the

11consequences thereof. Apart from these, there are support systems e.q. Station Electric Supply, Compressed Air which are active durinq normal operation and are also essential in the functioning of the ESFs.

Since process systems play an active role in plant operation, any process equipment failure would be immediately annunciated. But in case of protective and containment systems, being normally standby, there may be component failures which will be unrevealed till there is a demand on the system to function or it is tested. As a result a safety system will remain in a failed condition over the period of time from the occurrence of the failure till it is revealed by the test and repairs are effected. A process system failure during this interval would result in a dual failure. Thus, an accident sequence would arise if a process failure is coupled with the unavailability of one or more ESFs.

3,1,.RELIABILITY-CRITERIABased on the system definitions above, the reliability

index of process systems or lEs has been computed in terms of frequency i.e. the probable number of failures per year while for the safety systems, the. term unavailability or probability of failure on demand has been used which is the probable fraction of the time during which the system is not available. The unavailability is further related to component failure rates and test frequencies by the following equation

UnavailabilitysFailure rate(yr_1)*Failure- duration(Yrs)

where the failure duration is assumed to be equal to half of the time between tests since the failure at any time between tests is equally probable. Unless mentioned otherwise, the test interval used in the analysis is assumed as one month. Small variations in the test intervals are not considered whereas, if it varies by an order or more, an exact computation is used. In addition, the contributions due to scheduled and breakdown maintenances are also incorporated. The distribution of downtime is assumed as lognormal, with a median duration of

24 hours and a maintenance action rate of once in six months.

3.Jg.IMbPRE RATE DATAThe input data required for reliability analysis

comprises of the following

i) Component Failure Rate Data

ii) Component Maintenance Data

iii) Human Error Rate(HER) Data

iv) Common Cause Failure(CCF) DATA

The confidence in reliability analysis is determined to a large extent by the accuracy in failure rate data of the constituent components. It would be ideal to use data based on our operational experience but this is presently not adequate. The other alternative is to use data from established sources[4] which may not be always applicable due to variations in design, quality, operating environment etc. Bayesian techniques have been used to obtain better estimates by using the limited information based on RAPS experience and WASH-1400 as prior for a number of components like DCs, Transformers etc. The Kolmogorov- Smirnov test of hypothesis applied to the posterior confirmed its lognormal distribution. The Bayesian analysis of DCs is shown in Table 1.

3.3-C. CAPSE. EAIMRESThe common cause failures are dependent, multiple

failures arising from a common initiating cause. The main categories of CCFs considered in the analysis are

i) Design Errors

ii) Manufacturing Errors

iii) Test and Maintenance Errors13

iv) Effect of External Environnent

As far as practicable, care is exercised to 3se«p the process and safety systens independent of each other and safety systens anong themselves to nininise the incidence of CCFs. Special qualification procedures where applicable, have been adopted for the components to viths-4nd the connon causes such as earthquake, accelerated environment following an accident like LOCA etc. $-factor rtel has been used for the analysis of CCFs and the plant specifics have been incorporated.

3.4 SAFETY AMD-RELIABILITY ANALYSIS

Fault Tree Analysis has been extensively used and safety and reliability analysis of various IEs and ESFs applicable to NAPP has been performed to obtain both1 the probabilities of failure on demand as well as spurious failure rates. This helped in the design evaluations and also, ill decision making regarding safety issues. Sone design modifications as a result of the analysis are outlined here.

i) Reliability inprovenent in the design of interlocks and D,0 condensate lines in the Reactor Building Isolation System to effectively isolate the containment.

ii) Design modifications in ECCS to account for the interdependence of various stages of injection,identification of components to assist in improved procurement and test procedures

iii) Comparative evaluation of designs for secondaryshutdown system to obtain optimum configuration from the viewpoint of simplified design resulting in high availability, reduced test and maintenance efforts and adequate safety

iv) Provision of isolating valves in the interface of moderator circulation system with liquid poison addition systems to reduce the frequency of loss of moderator

These modifications( Table 2), interalia, have been implemented to improve the system reliabilities contributing to an overall risk reduction. The results of reliability analysis of the various ESFs are ahown in Table ЗА. The frequency of failure in respect of various IEs is shown in Table 3B. In order to account for the variability in data, an Uncertainty Analysis has been carried out assuming a lognormal distribution for the failure data and error factors as listed in [4]. The details of Fault Trees and other calculations of the reliability analysis are given [5] and these are also included in the appendix. A distribution is obtained for the Top Event using a Computer Program based on Monte Carlo simulation and the 5th, 50th(median) and 95-i:h percentile points are obtained, for various reliability indices. The data on HER and human responses to accident situations is inadequate in our context. However, components prone to human errors and their effect on system functioning have been identified in the analysis e.g. the valves in Feed Vater System.

4 ACCIDENT SEQUENCE IDENTIFICATION

In view of the 'Defense in Depth* approach applied in the design of reactor systems, an accident situation arises only when an IE is coupled with the unavailability of one or more ESFs. Thus a dual or multiple failure is necessary for an accident to occur. These dual or multiple failures are known as Accident Sequences in PSA parlance. The significance of accident sequences can be understood from the definition of risk as follows:

Risk *• Probability of occurrence*Consequences

In a HPP, the probability of occurrence signifies the probability of all the accident sequences and the consequences are measured in terms of radioactivity releases. Thus risk from a NPP

IS

• IProbability of accident sequence*Consequences All accident sequencesand the overall risk can be quantified if we can identify all the accident sequences and evaluate their consequences. In level I PSA, the requirement is to identify all the accident sequences and relate them to component failures and human errors. In the present study, accident sequences relevent to NAPP have been identified' using Event Tree methodology. Event Trees for all the dominating IEs have been drawn the details of which are given in the following sections.

4 1 ACCIDENT..SE •10 > iCE-QBABIIFICAIIQN

The accident sequence as identified by the Event Tree may be expressed as follows.

; Accident Sequence-Initiating Event * ESF(s) Failure

Obviously, in an accident sequence there are other terms implying the success of other systems; however these

i can be ignored since the success probabilities are approximately 1.0. In terms of probabilities, the accident sequence probability may be written as

P ” PIE*PESF1*PESF2*.....

where PIEis the frequency of the Initiating Event and p_ is the probability of failure on demand or the unavailability of that particular ESF obtained from the respective Fault Trees. In order to obtain correct accident sequence probability, the correct probabilities of the individual factors must be used, incorporating any dependency among the factors. Thus various system probabilities are treated as conditional probabilities and expressed as

PESF1 " PESF1/IE and PESF2 " PESF2/ESF1.IE

where PESFi/ie denotes the probability of ESF1 failure given that the initiating évent has occurred and so on. A simple multiplication of the probabilities can only be used

when the various factors are independent. The dependencies, if any, are included in the discussion on the individual Event Trees.

4.2 L0SS QFCQQLANT ACCIDENT(LOCA1 JEELS.

The different locations in a PHWR PBT piping where XiOCAs can occur are shown in figure 1. Unlike in PVRs and BWRs the diameter of the largest piping in PHWRs is much smaller, thereby limiting the radioactivity discharge rate in case of LOCAs. The coolant activity discharged into the containment is smaller due to the smaller PBT inventory in PBVRs. Depending upon the ESFs required to act upon, LOCAs can be divided into

1. Large LOCA-e.g. PBT header rupture

2. Medium LOCA'-e.g Endfitting failure, Feeder rupture etc.

3. Small LOCA- Instrument tube rupture, SG tube rupture etc.

Small LOCAs(the break area ~<0.1\ of 2A, A being the area of the largest diameter piping) handling is within the capability of pressurising pimps to start with and depending upon the storage tank very .low level( which. definitely is indicated) small LOCA handling systems can take care of the situation. For breaks at some locations like pressure tube and steam generator tube, the recirculation phase of small LOCAs may not be actuated as no water gets collected in the FM vault. This may lead to ECCS injection/recircUlation. Bowever, there is sufficient time available for the operator to take action and manually control the course of the initiating event. If there are clad damages, these are limited and significant release of activity is not expected. The various scenarios that follow depending upon the break location and the progress of the IE depending upon the operation of the ESFs is discussed in detail in the subsequent pages.

Large LOCAs are characterised by break areas greater than 10\ of 2A. These lead to fast depressurisation of the PBT which leads to subsequent ECCS injection and

recirculation. Because of the speed with which the IE propagates, operator actions are not expected/anticipated and accordingly all the ESFs that have to be operated axe designed to cut in automatically. Because of the fast depressurisation and subsequent low PHT pressure ECCS cuts in and continues to provide cooling such that core damage if at all is limited. The course of the IE with the associated ESFs is discussed in detail in the later pages.

Medium LOCAs (figure 4),break area between say 0.1\ of 2A and ~10\ of 2A, are characterised by a slow rate of depressurisation of the PHT system, especially in the lower end of the break sizes of the medium LOCA breaksize spectrum. Because of this the reactor remains at full power after the LOCA occurs for a sufficient period(**1 to 2 minutes) adding a good amount of thermal energy to the system. In addition, the ECCS flow may not be enough, during the initial phases and a sustained low flow condition(stagnation) in the fuel channels is expected which can lead to singificant clad damage. (In case Reactor Building does not get pressurised to initiate crash cooling of the boilers to allow continued ECCS flow through the core,(it is essential that depressurisation of the PHT system by blowing of the secondary side of the boilers be carried out manually by the operator) If this operator action is not carried out there cçn be significant clad damage. In NAPP a provision has been made to convert this type of situation into large LOCA by opening a pair of parallel valves during the light water injection phase.

4.2.1 LARGE BREAK LOCA

The ET for the initiating event large LOCA is as shown in figure 2. It is important to note that the void coefficent of reactivity is positive in a PHWR and this warrants a fast shutdown in the present case. Since the moderator is not dumped on a reactor trip, the presence of a large volume of moderator which is cooled by an independent circuit of pumps and heat exchangers acts as an ultimate heat sink. Various studies [6],[7] indicate that no fuel melting is likely to occur even il ECCS fails on LOCA. Thus fuel melting in a PHWR can be postulated to occur when there is a breach in the moderator circuit in

conjunction with LOCA and ECI failure. Thus, the probability of failure in caee of accident sequences 2 and 7 would be further multiplied by the probability of loss of moderator. The other accident sequences are related to the failure of containment functions e.g. reactor building isolation« reactor building cooling etc. It is recognised that RBI system is extremely important in case of activity leaks and as mentioned in section 3« care <.«s been exercised in engineering reliability into the system. Further, double containment is provided to check activity leakages. RBC function is performed by i) Suppression Pool which is a passive system and ii) Fan Cooling Units. In Indian PHWRs, all high enthalpy systems including PHT are located in.a volume called V1 which is connected to rest of the volume V2 of the containment by means of a vent system via the suppression pool water. Any leakage from volume V to V2 which are seperated by leak tight walls and floors may marginally affect the efficacy of suppression pool, for which the probability is low. The vapour suppression pool would absorb about 25 to 30% of the energy released from PET system and also trap a significant part of the activity. The RB coolers located within the containment bring down the pressure following the accident.

.SMALL..BREAK LOCA •Small Break LOCA, as mentioned before, is defined as

the break corresponding to upto 0.1% of the double ended or about 1/2* in the PHT, which is more likely due to an instrumentation tubing etc. and is within the Pressurising Pump capability. Here, consideration is given only to those breaks which result in spillages in FM Vault/Boiler Room area. Steam generator tube breaks and pressure tube ruptures are considered seperately. Due to the high stored energy and significant decay heat for the first few minutes, it is essential that pressurising pumps operate for about half an hour of the event.. Since these pumps are on class IV, the failure of class IV supply in this duration would affect the core cooling capabilities. Even though the FM pumps start in case of class IV failure, the flow delivered by these pumps would not be adequate to meet the core cooling requirements, resulting in a significant rise in clad temperature. Since the system goes on loosing

inventory, the pressure falls and ECCS would be actuated eventually. But till such time, due to inadequate coolinq, some clad damage might have already occurred.

However, a Small LOCA Handling System(SLHS) has been provided in NAPP wherein, on sensing a low level in the storage tank and at PHT pressure greater than 55Kg/cm2, D20 is injected into the PHT storage tank from the ECCS accumulators and subsequently, the spillage is recirculated to allow enough time for operator action to detect and plug the leak. Even with class IV power available, failure of SLHS is also likely to lead to the same situation as described above. However, if small LOCA injection occurs and recirculation fails, the situation is less severe as the stored heat from the fuel has been removed and the clad temperatures are low. In this situation, high pressure ECCS injection is not possible as the D20 in the accummulators got already transferred to the storage tank. The efficacy of SLHS in situations where a small break may propagate into a medium break is also questionable since’ the accummulator water would not be available for emergency coolant injection. It may be worthwhile under the circumstances to transfer initially D20 from outside tank and subsequently from the D20 accummulators. This would ensure availability of ECCS during injection phase. The event tree for small break LOCA is shown in figure 3.

4.2.3 STEAM GENERATOR TUBE RUPTORE

Upon the rupture of a steam generator tube, there will be an increase in the affected SG water level with a decrease in the D20 storage tank level. Unlike in RAPS and MAPS boiler level control action is on the individual boiler level and steam flow rate from the boiler. The level controller, thus, tries to maintain the boiler level. However, as the PHT system is loosing water, the storage tank level dips to very low value and this initiates a reactor trip.

As the system pressure is greater than 55Kg/cm2 and D20 storage tank level is low, small leak handling system gets actuated. The D20 required for this system is drawn initially from the ECCS D20 accumulators. When the level in

these is low, a D20 storage tank outside the reactor building provides the required makeup water. This can continue until the D20 in the 3211-TK-1 gets exhausted.Before this water is exhausted, the leaky SG should be identified and isolated. The positive indication for SG tube leak is high activity in feed water(or steam). As the steam activity is continuously monitored, it is possible, by operator action, to isolate the leaky steam generator.

If the leaky SG is not promptly isolated, the PHT system starts loosing water and hence pressure. ECCS accumulators, being empty, are of no use. Due to reduced cooling, the clad temperatures start increasing(due to decay heat). Eventually light water injection will start and stabilize the system with light water recirculation. This IE is thus, more or less identical to the small break LOCA in terms of the effects on the core and PHT but may lead to a large scale contamination in the turbine building etc.

4L-2 ^-PRESSURE. TUBE ^CORRESPONDING CALANDRIA TUBE RUPTURE

The PHWR comprises a large number(306 at NAPP) of pressure tubes, each about 5.43 meters long having 83mm diameter and each is surrounded by a concentric calandria tube. A large inventory thus enhances the probability of failure and as described in figure 1, this would be a case of LOCA inside the core. Based on the operating experience of Canadian and Indian PHVRs, a failure rate < 1.0*10’4/year per pressure tube is obtained(which is corroborated by the failure data usually used in the pipe rupture calculations). With respect to the break size, effects on the PHT system, the pressure tube failure is equivalent to a medium LOCA. However, a significant difference exists in the present case since the failure may induce large reactivity effects due to dilution of borated moderator with PHT heavy water in addition to the positive reactivity effects due to crash cooling, voiding etc. The fast reactivity changes would be compensated by the simultaneous actuation of both the primary and secondary shutdown systems and the slower dilution of moderator can be overcome if the Automatic Liquid Poison Addition

System(ALPAS) is effective in injecting the boron into the moderator. The efficacy of ALPAS in the prevailing conditions needs be confirmed.

4-.-2_. 5-MAIN STEAM LINE BREAK

This IE is somewhat identical to the Design Basis Accident(DBA-Large LOCA). However« the energy released into the containment, the subsequent pressure peak and temperature rise would all be in the secondary containment which is not designed for pressure retention and this is vented to the atmosphere by the opening of the blowout panels. The probability of this IE is also expected to be greater than large break LOCA. The effect on PHT system in terms of depressurisation would be very fast, due to crash cooling but the pressure would be restored if there is no leakage in the primary system through SG tubes. The event tree for steam line break without any steam generator tube failures is shown in figure 5A. Boiler inventory would also deplete fast due to crash cooling and Fire Water System(FFS) would have to be actuated by the operator when the Boiler pressure falls to about 3.7Kg/cm2(Both the boiler feed pumps may trip due to overspeeding and AFWS may not be able to provide adequate cooling). The reactor cooling may continue in this mode or SDC system may be valved in. In case FFS is not available, a fast operator action is warranted to bring in SDC system Within about 20 minutes before the boilers dryout. If SDC fails under these circumstances, it would result in core damage. In case ECCS fails, 20-25% voiding is expected resulting in an uncertainty in the effectiveness of the thermosyphoning process to cool the core and may cause clad.damage. If FFS also fails, a large scale clad damage is expected though the presence of moderator may prevent a core melt.

The reaction forces coupled with a high pressure differential across the SG tubes, may induce multiple failures in SG tubes and the event tree for this is as detailed in figure 5. Thus MSLB may result in medium/large LOCA too and it is essential to qualify the SG tubes for such impact loading.

4.3 FEED WATER SYSTEM РАТТ.ПИЕ

Thé event tree for this IE ie shown in figure 6.Feed Water System failure can be due to a) failure of boiler feed pumps(BFPs) or b) rupture of feed pipe. IN case of failure of BFPs reactor trips on high PHT pressure or low steam generator level. Auxiliary BFPs start on auto. Reactor can be cooled down further, if it is warranted, by the SDC pumps.

Feed Water pipe break downstream of the check valves near the SGs leads to a situation where the break can neither be isolated nor auxiliary feed to the SGs be supplied. This leads to a complete blowdown of the affected SG. This is likely to induce SG tube failures due to thermal shock as well as loading of the tubes due to vapor bubble formation and collapse on the secondary side of.SG . However, the length of the piping( and hence the probability) that results in such failures is small. This SG can, by operator action, be isolated on the primary side(due to continuous monitoring of steam activity in the individual boilers).

Feed water pipe breaks upstream of the check valves near the SGs can be isolated(The check valve acts as an isolater on one side of the break). Even if the break cannot be isolated from the otherside this situation does not lead to loss of SG inventory. If the operator valves in the auxiliary feed line, normal cooldown of the PHT can be resorted to and later SDC system can be valved in. If the auxiliary feed line cannot be valved in, secondary steam relief is required and SDC system has to be valved in for entering into a stable state. In case SDC system cannot be valved in Fire Water injection to Boilers is required. If Fire Water System also fails, cooldown of PHT is not possible.

If secondary steam relief is not realised, PHT gets pressurised and this will lead to either PRV opening (and hence LOCA) or pressure tube rupture.

1...4 CLASS IV-POWER SUPPLY FAILUREClass IV is the nain supply provided both by the grid

and generated by the station. This IE is significant in our context due to high frequency of class IV failure. Based on the operating experience, it is observed that the frequency is > 1.0/year which is relatively high since it is usually 0.1 to 0.3/year in many other countries. Interdependence( common cause failure) of station supply on grid fluctuations and vice-versa is a significant factor in determining the frequency. At NAPP in the Northern Grid, the frequency may be of the order of 2/year.

The ET and the various ESFs required to mitigate the effects of the transient are as shown in figure 7. The Secondary Steam Relief(SSR) is provided by a redundant configuration comprising steam discharge valves(SDVs) and boiler relief valves and the probability of failure of this system would be low. On class IV failure, secondary cooling is provided by auxiliary feed water system which is further backed up by the FFS driven by three dedicated diesel pumps. In case of loss of all secondary cooling, SG holdup would last for about 20 minutes and by this time, the. shutdown cooling system must be valvèd in. Similar criteria is applicable to valving in of FFS in case AFWS is not available.

Accident sequence 14( fig. 7)is critical which depicts the failure of both class IV and class III leading to the situation of Station Blackout. The station batteries are usually rated for a duration of 20 to 30 minutes and this sets a limit to the available time within which the power supply must be restored. The probability of restoring the supplies in 30 minutes is low. NUREG-1032 of USNRC quotes a median value of restoring time for offsite power as 0.5 hours and repairing time of DG as 8 hours. In case of an extended Blackout, it would result in a critical situation since AFWS, SDC would not be available. Also, the supply to control and protective systems(class I) would be lost, resulting in a total loss of minitoring and indication of the plant status. It may be essential to crash cool the primary which would result in large scale

voiding in the system. However, with secondary cooling available, provided by the PFS, thermosyphoning may be effective. The reliability of FFS is thus crucial for mitigating the station blackout situation. In addition, FFS is essentially a low pressure system and manual actions are involved in valving in of the same. In case of a station blackout, which is definitely an unusual situation,the stress on the operator is likely to be high and the time available is also "half an hour. Hence the probability of human error would be significant and the same is considered as 10’2 . However, sicne FFS is the only safety system available, the chances of recovery would be high and this system could be valved in after some delay.

A ._5_- COMPRES SEP AIR SUPPLY SYSTEM FAILURES

The compressed air is an important support system required in instrumentation and control for valve operations, pneumatic process monitoring components etc. The system comprises mainly of compressors and dryers and dry air is supplied to a common header through air receivers which act also as air reservoirs for 5 to 10 minutes. In the reactor building, three air receivers have been provided to provide triplicated lines for air supply to safety system components. These air receivers further act as reservoirs and in case of losa of compressed air supply from the compressors, the capacity would be available for 10 to 15 minutes. Utmost care is exercised in safety system design wherein, the air operated devices would assume fail safe operation and no continuous air supply would be required for further operation. In case where failure of compressed air may lead to unsafe situations, local air receivers have been provided to mitigate the sitiation. Thus, the reliability of compressed air system would be adequate if the local air receivers are properly maintained so that they act as proper standby components. It would be essential to provide pressure gauges in the local air receivers and institute periodic maintenance. At this level of operation, compressed air system failures may not affect the reactor safety.

A. fi- PPEL HAUDLING ГА1ЫтЕа

During refuelling operations! «Fuelling Machine is a part of the PHT systea and foras an extended boundary. Various failure nodes arising froa the FM operations are considered as followsa). Fuelling Machine On Reactor

In this operation, there is hydraulic connection between the nagazine and the reactor end fitting. All the DjO leaving the aagazine has to return to the priaary systea through the coolant channel. Thus, any breach in the FM D20 circuit would result in LOCA. However, based on the reliability analysis, the contribution of this interface LOCA is negligible,b) Fuelling Machine Park

In this node, FM is in the transit between the reactor and the fuel transfer port and could be carrying spent fuel bundles in the aagazine. A failure in the D20 control systea would result in spent fuel bundles in the aagazine not getting cooled.

The other possibility nay be that the seal plug and sheild plugs are.not seated properly after the refuelling operation and the FM is renoved froa the channel after the leak detection test fails to diagnose the leak. This aay result in the ejection of seal and shield plug leading to endfitting failure and ejection of áll the fuel bundles froa the channel.c).On Fuel Transfer

Xn this node the FM is claaped to the fuel transport and there is a hydraulic connection between the aachine and the fuel transfer port. Failure of D20 control systea would result in fuel bundles being deprived of the cooling.

The frequency of fuel handling failure is estiaated as 1.0*10°/year resulting in radioactivity leaks into the containaent. This IE would be significant only when the RBI systea fails. Thus, the probability of this accident sequence is less than 1.0*1 O'*8/year.

■4.? LOSS OF REGULATION

26

The regulating system is designed to control the reactivity changes and thereby, providing a control on reactor power and preventing any flux peaking etc. The loss of regulation Accident(LORA) amounts to withdrawl of control and absorber rods used for reactivity control. The regulating system is based on triplicated instrumentation channels and the movement of any rod is controlled by the individual servo motors. Thus failure of the servo motor would affect a single rod only. In case of a failure in any regulating channel, the same is rejected and the control is transferred to an operating healthy channel. Thus, a single channel faulure coupled with the failure of the transfer circuit would result in loss of regulation in a single control channel amounting to about 5mK. In case of the failure in all the regulating channels, due to a ■ common cause, it is a LORA involving about 16mX reactivity ; insertion. However, the design of the reactor ' regulating system limits the reactivity addition rate to a maximum of 0.03mK/sec per control group. Thus, in all cases of LORA involving control and absorber rods, the maximum rate of reactivity insertion would be limited to 0.06mK/sec. Also, each of the two reactor shutdown Systems have enough depth and speed to terminate all reactivity insertions. However, reacticvity insertions would lead to high pressure in the PHT system. If the pressure relief system is not available, the situation may lead to pressure tube and calandria tube: failure. This would result in the crash cooling of PHT system. Positive reactivity introduced may exceed the worth of the shutdown systems and the efficacy of ALPAS would also be questionable. The frequency of this accident sequence would, however, be "I.0*10’*/year.

4.8 FLOW BLOCKAGE IlLAHY. CQPLAHT-ASSEMBLY.

Operating heat fluxes, even in maximum rated channel, are such that there is sufficient burnout margin. In fact flows upto ”18\ of the rated flows[5] in the channels can be tolerated without any fuel damage. Bence flow reductions even of sufficient magnitude do not pose any problem.

Ofcourse these result in high channel teaperature «rich causes a setback of the reactor. Thus only flow blockages( or severe flow reductions) are of concern. These can be due to suspended objects of 'considerable size in the PHT system. Normally objects left in the PHT system during construction can be detected during commissioning. Water chemistry in PHWRs is well controlled and this minimises corrosion and thus buildup of suspended particles. During refuelling operations, some paddings etc. can get dislodged from the fuel bundle. These are unlikely to lead to full channel blockage. Broken parts from rotating components, like pumps, can also cause some kind of flow blockage. Similarly parts of valve bodies, which if they get dislodged and get deposited elsewhere,, can be a cause for flow blockage. However in the operating PHWRs in the world, there is not even a single event of this type reported so far(eventhough partial blockages have been reported). If flow blockage occurs, there is no safety system to take care of. However, moderator can act as a heat sink and the damage is limited to the affected channel only.

4.9 ACTIVE. PROCESS..WATER SYSTEMAPW system failures of short duration are unlikely to

have any impact on the safety of the reactor. In case of failures of long duration, the reactor automatically trips as this system affects the cooling of thd PHT pump motor, moderator pump motor etc. If emergency APW system gets valved in, moderator cooling can be maintained.

To keep the PHT system solid( to compensate for the shrinkages) the pressurising pumps have to cut in and continue to operate while the auxiliary feed water system continues to cool the PHT system.If the pressurising pumps do not cut in, D O injection of ECCS can occur. If AFWS does not cool the primary, this might lead to high PHT pressure and consequent IRV opening which if fails to close again is a LOCA situation.

In case emergency APW system fails, moderator cooling is affected, the remaining accident scenario remains unaltered i.e. AFWS operation to remove the decay heat, pressurising pumps/D^O injection to keep the PHT solid. If

all these systems operated as intended, the situation is stable. If the pressurising pumps and the DO injection fail to come, there may be voiding in PHT system and if the voiding is substantial clad failures may occur. In case auxiliary feed water system does not come in, PHT gets pressurised and this leads to IRV opening and hence LOCA. The event tree for this IE is shown in figure 8.

Д.-ДО HOH~ACTIYE HIGH PRESSURE PROCESS WATER SYSTEMThe event tree for this IE is shown in figure 9. If

the loss of NAHPPW system is of short duration, it does not affect the safety of the reactor or its associated safety systems. If this failure is of significant duration, systems like DCs are affected due to the nonavailability of jacket cooling water system. Under this situation, class IV failure of significant duration will result in a condition similar to Station Blackout.

In case class IV is available, as the main and auxiliary boiler feed pump cooling is affected, secondary steam relief is required to depressurize the PHT system or else IRV may open, close, open again to relieve the excess PHT pressure and this may ultimately result in LOCA. If SSR is avilable, to keep the PHT solid, auxiliary main feed( pressurising pump operation) is required. In case this is not available, SDC can be valved in which if fails to cut in, is likely to lead, to fuel overheating and subsequent clad failures. If SDC cuts in, the consequences will be less. If auxiliary main feed is available, and SDC is available, a stable state is reached. If SDC does not come, fire water has to bè injected into the boiler to continue cooling of PHT. However, it is not certain whether enough : cooling of PHT system is provided and this may result in some clad failures. If fire water does not get injected, PHT gets pressurised and IRV opening may take place and this is going to be a delayed LOCA.

з.-SYSTEM DEPENDENCIESAs mentioned before, various ESFs have been designed

to operate independently- both among themselves and also, with respect to the IE. However, some form of dependency has been observed. Normally, it is expected that various components and equipments are designed to operate in the accelerated environmental conditions generated by the IE. In case of LOCA, an environment of high temperature, pressure, radiation and humidity prevails in the containment and various components e.g. pump seals, pump motors, junction boxes, coolers etc. are susceptible to it. Further, the presence of moderator as a heat sink is very important in case of PHWRs to prevent fuel failures' if ECCS fails büt the efficacy of the system need be ensured when a significant amount of energy is added into the moderator. The reliability of the moderator pumps, flange joints etc. will be affected in such cases. The effects of such common causes have been incorporated in the accident sequence quantification. In addition, intersystem dependencies have been considered in developing the event trees. Effect of the failure of the support systems on subsequent progression of the event trees have been incorporated( eg. DCs are not considered in NAHPPW system initiating event ET) where ever found necessary.

fi-PQUIHAXINfi, ACCIDENT.SEQUENCESThe overall number of accident sequences identified

through Event Tree analysis is very large as described in the previous section. However, based on the probabilistic and analytical assessment of the consequences of the accident sequences; a relatively small number of accident sequences which contribute to activity release in to the containment are presented in the table 4. These accident sequences( excluding those resulting in coolant activity only i.e. sequences 2,3,16 and 19) are expected to result in core damage and are designated as Dominating Accident Sequences. The extent of core damage is not assessed at this stage and the same will be the next phase of this study wherein, the consequences in terms of the effect on the core, radioactivity released, effect on the containment and its failure modes etc. would be discussed. As mentioned

before, the presence of moderator in the core prevents fuel eelting in case of LOCA and unavailability of the emergency core cooling systems. Thus, all accident sequences originating from LOCAs, MSLB and others resulting ultimately in LOCA would cause fuel failures only if the moderator is not available as a heat sink. Thus, accident sequences resulting form station blackout(9 and 10) and from active process water failure which provides cooling to moderator heat exchangers, contribute to fuel failures as such and in all other cases, the frequencies would be multiplied by the probability of loss of moderator as a heat sink. The frequency of core damage, thus calculated, is 1.5*10'S/year. The major contrubution is from station blackout related events( A decision has been taken to incorporate a third DC which reduces the emergency power supply unavailability by a factor of 3, leading to a core damage frequency of '5*10"e/year). An uncertainty analysis has been carried out for all the dominating accident sequences( table 5) and the error factors are further propagated to obtain a distribution for the frequency of the core damage as shown in the table 6.

7 ACCIDENT- SEQUENCES-SENSITIVE TQ..HUMAN ERRORThe boiler feed water system is backed up by the AFWS

which automatically cuts in when main feed pumps trip or class IV failure occurs. As a result of TMI studies and recommendations, emergency feed water is provided to boilers from the fire fighting system. This is essentially a low pressure system and thus, would necessitate a crash cooling on the secondary side to reduce the pressure and prevent a SG dryout before fire water is injected into the boilers. Thus, in nccident sequences involving loss of main and auxiliary feed water, a reasonably fast operator action(20 t 30 mts) is warranted to provide secondary cooling. Funner, SDC too is a low temperature system and is reliable if valved in when PHT temperature is <150°C. In case all the secondary cooling is impaired, a crash cooldown of the PHT is essential to valve in SDC. This is a high stress situation and the available time is short and the information available to the operator also may not be adequate resulting in a high human error probability. In fact, so far, training is with respect to failures in

various process systens only which oust be extended to accident sequences.

8 relative importance of initiatingSYSTEMS

¡Vis i»цз^шр. mmAccident sequences are related to the IEs and the ESF

failures which are further expressed in terns of the component failures and the human errors. This provides a measure to reduce the core damage frequency. The dominating XEs considered for the analysis have been explained in section 2 and the ETs were constructed for these IEs to obtain all the accident sequences and then the dominating accident sequences. The contribution of the various IEs to the dominating accident sequences is shown in the table 7. In case of an ESF, the importance function is expressed in terms of the core damage frequency reduction worth and is defined as the ratio

IAir_dominatinq_jequences_not containing a particular ESFCA11 dominating accident sequences

which, in other terms, is a measure of the effect of the reliability improvement in the particular ESF as shown in the table 8.

The authors are indebted to Dr. R.N. Kulkarni of Atomic Energy Regulatory Board for his immense help in preparing the required computer programs fox carrying out the uncertainty evaluations. Thanks are also due to the various system designers at NFC for useful discussions. The various comments received on the draft version of this report from the organisations BARC,HPC and AERB were appreciated and have appropriately been considered in this version.

1. WASH-1400, Reactor Safety study, USNRC(1975)

2. Gernan Risk Study

3. Safety Series no.S0-SG-D11,Safety Guides -General Design Safety Principles for NPPs, IAEA, Vienna

4. National Reliability Evaluation Progran(NREP)<1982)

5. NAPP Safety Report Voluae XX*6. CANDO Safety Research -A Status Report, Second Annual

Conference, Canadian Nuclear Society, June,1981, Hancox, W.T.

7. Safety Research for CANDO Reactors, IAEA. Technical Connittee Meeting on Theraal Reactor Safety Research, Moscow, Dec,1981, Hancox, W.T.

8. Application of PSA in licensing of PHVR, Tareny E.M.,Presented at the XAEA-AERB workshop on Safety Analysis held at Bombay,5-16 May,1986.

TABLE It BAYESIAN ANALYSIS OF DEMAND FAULURE8 DATA ON DG8 AT RAPS

PRIOR POSTEROIR

Vos O.Ol 0.0433

A • _ 0.03 0.03330 SO

X • 0.09 0.6260 95

M -З.Э2 -2.93

0 0.673 0.943

SAFETY SYSTEM

AVERAGE FAILURE RATE/

UNAVAILABILITY

MAJORITYCONTRIBUTORS

CRITICAL COMPONENTS

AND CONTRIBUTIONREMARKS/

RECOMMENDATIONS

MODIFIEDVALUES

EMERGENCY CORE COOUNG SYSTEM

«x tÓ3/dLIGHT WATER. MX

4.5 X 1Ö3

RECIRCULATION17X1Ï3

RUPTURE DISC 2 * Ml MV-52 1 X mlMV-24 1 к Ml

MV-31 1 X ÎÔ

CHANGE M RO DESIGN PR. REDUNDANCY IN MVs. ENVRNMENTAL OUALFICATION OF MVsAND PIMPS.

2.5 xM3

REACTOR BUILDING ISOLATION

8.x 10/d BYPASS LINE4 X 10"3/ 4 CONDENSATE RETURN LINE4 X 10^/ d ;

DAMPERS 7314 0M5 ORDM6 - 2x10vd teach)VALVES-Ixl^/d (..chi

INTERLOCKING THEOPENING OF 734 DM-13 ANO OM-14 WITH THE aOSURE OF DM5 i DM«. REDUNDANCY M CONDENSATE RETURN UNES

<1xM3

ELECTRICAL POWERsupply ¡missii

Щ CLASSE1-2/yr.SxlïVd

CCF DUE TO OFSIGN,MAlNTENA4a FIÆL OH SYSTEM ETC

DGs - 3x1? (CCF) 'INDEPENDENCE AND STAGGARO MAINTENANCE.

3x1?

MODERATOR SYSTEM9 DROP N LEVEL ñ) CALENORIA

ISOLATION

2V.

9xW3/d

SHM t REG. ROO COOLMG SYSTEM ACTUATION CIRCUIT

LARGE NO. OF BELLOW SEAL VALVESMVs » 3 x lôV d (each)

MONITORING OF MV ACTUATION CIRCUIT COMPONENTS.

SxM3

REACTOR PROTECTION SYSTEM

INSTRUMENTEDREÜEF VALVES

2x10/d

<10/d

MSTRUMENT ATION CHANNELS CCF

INSTRUMENTATION CHANNELS CCF

REDUNDANCY M ТИР PARAMETER AM)SHUTDOWN DEVICES. .

?

TABLE SAFETY SYSTEMS MAIN FAILURE MODES & MODIFICATIONS

TABLE ЗАi FAILURE PROBABILITIES OF SAFETY SYSTEMS

8.No SAFETY SYSTEM DEMAND FAILUREPROBABILITY

ERRORFACTOR

1 Emergency coreCooling 3.5*10'3 2.0

2 Reactor BuildingIsolation 2.0*10'* ¡2.4

3 Emergency PowerSupply 3.0M0'3 1.Э

4 Reactor BuildingCoolers l.OMO'3 3.0

S Reactor Protection 2.0*10'* 4.06 Auxiliary Feed

Water 1.0*10'3 1.8 •7 Fire Fighting l.OMO"3 3.08 Small LOCA

Handling 2.0M0"2 3.09 Moderator System 3.0M0"3

TABLE ЭВ: FAILURE FREQUENCY OF INITIATING EVENTS

S .NO INITIATINGEVENT

FAILUREFREQUENCY(уr"1)

ERRORFACTOR

1 Large Break LOCA 1.0*10"4 102 Medium Break LOCA 1.0*10"2 33 Small Break LOCA 1.0*10"2 34 • Class-IV Power

Supply 1.0 25 Feed Water System 0.5 1.76 Main Steam Line

Break 3.0*10° 37 Active Process

Water System 1.0 1.58 Non-active High

Pressure Process Water System 0.2 2.1

9 Compressed AirSystem 2.6 1.3

10 Moderator Level 2.0 1.211 Steam Generator

Tube Rupture 1.6*10° 3

37TABLE 4i DOMINANT ACCIDENT SEQUENCES - CONSEQUENCES

S.NO ACCIDENT SEQUENCE FREQUENCY(умг"1)

CONSEQUENCES/REMARKS ¡IE E8Fe(Fa1L«d>

1 ML ECR 2.0*10"5 Clad failures expected. May lead to delayed RB pressuri­sation depending up on the extent of MU reaction.

2 ML RBC 1.0*10’S RB remains pressurised, high leak rates expected. No core damage

3 ML . RBI 1.0*10 5 All coolant activity released to environment. No core damage

4 ML ECI 1.5*10*5 Significant clad damage expect -ed leading to MU reaction &H generation resulting in contain -ment pressurisation. Fuel melting not expected due to presence of moderator.

S SL SLHSI 1.0*10*4 Uill lead to clad failures.MU expected when ECI comes in contact with clad at high T.H generated may lead to conr tSinment pressurisation.

b SL RT 1.0»10*e Even SLHS may not be available since the same signal is used for both. Significant clad damage(Studles required)

7 SL CLASS-IV 1.0*10*5A

Voiding occurs. SignificantClad failures expected.

в CLASSIV

CLASS'-! 11 <1.0*10"3 If exceeds 0.5 hours results in clad damage.Recovery of off site power considered here.

9 CLASSIV

CLASS-11IfFF8 1.0*10*®

m

PHT pressure singal not avai­lable. Spring loaded RVs open. Otherwise pr. tube rupture. NO ESFs effective to cool the core hence core melting likely.

10 CLASS CLASS-1II,HE 1.0*10*' Same as above11 CLASS

IVCLASS-1II|SSR <1.0*10*7 Same consequences as above.

12 CLASSIV

SSR <1.0*10*4 May result in №.. All otherESFs including moderator are available. It will result in significant clad damage.

TABLE 4 continued fron previous page

S.NO ACCIDENT SEQUENCE FREQUENCY(year’1) CONSEQUENCES/REMARKS

IE E8Fs(Fa1Led)13 FW 88R <3.0*10*5 ML. All ESFo Including aodera-

tor systea are available.14 FW AFUS &

HE(10‘2)3.0*10’e Saae as above.

13 FW AFWSf8DC i HE(0.9> 4.3O10’7 ML,

But ECC8 nay not be effective since crash cooling is not available.

16 MSLB 3.0*10’3 All coolant activity released Into the sec. containment. High leak rates expected.

17 MSLB ECR 6.0*10 e Clad damage evulvalent to ML/LL depending upon the number of tubes falling. All activityInto the sec. containment.

18 MSLB EC1 4.3*10’® Significant clad damage.19 MSLB FF3 & HE 3.0*10’7 LOCA through 1RV. Only coolant

activity released.20 APU EAPWS 1.0*10 * Some clad damage may occur.21 APW EAPWB & EC1 1.0*10‘7 Clad damage, heavy voiding of

PHT expected.22 APU EAPWS i AFUS 1.0*10 7 ML. No moderator cooling.23 NA

HPPW8 CLASB-1V2.0*10’* Same as station blackout.

24 NAHPPWS SSR

<2.0*10’® ML through 1RV. No crash cooling since SSR Is not available.

23 NAHPPWS SDC & FF8

2.0*10’7 LOCA through IRV

TABLE S» DOMINANT ACCIDENT SEQUENCES - UNCERTAINTY ANALYSIS

8.No Accident Sequence Medianrrequency<Yr"1)

ErrorFactor

1 ML & ECR 6.0*10"® 3.34 ML & ECI 4.3*10"® 3.35 SL & SLHI 3.0*10"7 3.66 SL & RT 3.0*10"® 4.7'7 SL & Class IV 3.0*10"® 4.7В Station Blackout 9.0*10"7 2.29 Blackout & FF8 1.0*10"® 3.9

11 Blackout & SSR 1.0*10"7 6.012 Class IV & SSR 3.0*10"7 3.713 FW & SSR 1.3*10"7 3.714 FWyAFWS & HE(.Ol) 1.3*10"® 3.4IS FW,AFWS,SDC S.HEC.9) 1.4*10"® 3.417 MSLB & ECR l.B*10"® 3.318 MSLB & ECI 1.4*10"® 3.320 APW & EAPW8 3.0*10"® 3.021 APW,EAPWS & ECI 3.0*10"10 3.622 APW,EAPWS & AFW 3.0*10"® 4.823 NAHPPWS & Class IV 6.0*10"7 3.624 NAHPPWS & SSR 6.0*10"® 3.723 NAHPPWS,8DC & FFS 6.0*10"10 3.4

* This corresponds to extended station blackout

\ 40\

TABLE 61 PERCENTILES OF THE CORE DAMAGE FREQUENCY

PERCENTAGE FREQUENCY

3 7.01*10"®10 8.21*10"®13 9.19*10"®20 1.00*10"*23 1.09*10’*30 1.17*10"*33 1.2DM0"*40 1.33*10"*43 1.43*10"*30 1.34*10"*33 1.66*10"*60 1.80*10"*63 1.93*10"*70 2.11*10"*73 2.30*10"*80 2.37*10"*83 2.93*10"*90 3.47*10"*93 4.48*10"*

TABLE 7 s COHTIRBUTIOH OF IES TO DOMINATING ACCIDENTSEQUENCES

S.NO IE \Contribution

1 ML 3.442 FW 5.453 MSLB 1.034 SL 10.925 APW 2.636 NAHPPW 1.207 CLASS-IV 75.11

TABLE 8: CORE DAMAGE FREQUENCY REDUCTION VORTH OF ESF

S.NO ESF CMFRW

1 SSR 0.822 SLHI 0.903 ECR 0.974 ECI 0.985 EAPWS 0.976 RT 0.9997 AFWS 0.9788 FFS 0.679 SDC 0.99910 CLASS-IV 0.9811 Mod. System 0.40

LOCA IESim raras

INSIDETHECORE

Pressure TubeRuptures

1INLET OUTLETSIDE SIDE

Inlet Header Break Pimp outlet Pipe BreakPuap Suction Piping BreakInlet Feeder Rupture

- Outlet-Header Break- Boiler Inlet Pipe Break- Outlet Feeder Pipe Break

FIG 1: VARIOUS TYPES OF LOCA mrmiHG EVENTS IN PHNRS

OUTSIDETHECOBS

ZUPHT

INTERFACE

Bleed Valves Stuck Open IKV OpenBreak in Bleed System Piping Break in Relief System Piping Feed System Pipe Break Shutdown Cooling System Pipe Break Gland System Pipe Break Fuelling Hachine Interface Failure

IE RT EP ECI RBI RBC ECR

23

45

67

8 9

1011

12*10"Z/yr

1*10'В/уг

2*10’’°/yr

1*10'7/yr

2e1O",0/yr

1.5*10 /уг

1.5*10'®Ууг

1.5*10*’,,/уг <10'? /уг

410 /уг

FIG 2: EVENT TREE FOLLOWING LARGE BREAK LOCA

IE PS RT SLHSI SLHSR

0‘*/yr

--------- 1

Class-IV

1

23

45

6

fWj/yr1*10 /уг

2*10"*/yr1*10's/yr

I^IO'^/yr

SLHSI Snail LOCA Handling Systen Injection SLHSR Small LOCA Handling Systea Recirculation

FIG 3 : EVENT TREE F0LL0MIN6 SMALL BREAK LOCA

IE RT 1 £P ECI RBI RBC 11 BCR

101112

2*10“*/уг 1*10" /уг2М0_;/уг1*10*5/уг2*10"*/уг 1.5*10"*/уг1•S*10"*/yr 1.5*10’5/уг1.5*10"J/yr

<Ю"1 /уг <10 * /уг

Пв 4: EVENT TREE ПЗШМ1Ж MEDIUM BREAK LOCA

3*10°/yr

6*10’*/уг 3«10"*/yr

6*10’9/уг

4.5*10*-/yr

ECÏ Eaezgency Cooling Injection RB? Reactor Building IsolationECR Energency Cooling Recirculation

FIG 5: EVENT TREE FOLUMNG NAIM STEM! UNE BREAK INDUCING STEAK GENERATOR TUBE FAILURE

1

3«10°yr10’*

1.5*10’3

10-2

10,-3 4ю-1

T24 3.0*10 e/jr5 3.0*10’?/yr6 3.0*10’I/yr7 3.0*10’Г/тг8 4.5*10*"/yr9 4.5*10’J/yr10 4.5*10’"/yr11 4.5*10"*Vy*121314

4.5*10 ¡J/yr 4.5*10 « /ух 4.5*10’B/yr

FIG 54: EVENT TREE F0LUM1NG MAIN STEM LINE BREAK (WITH N0 STEM GBCRÀT0R TUBE FAILURE)

SSR 1 AFNS HE 1 « 1 HE FFS J

BE Huaan Error

FIG G: EVENT TREE РОШЯПЖ FEED MATER SYSTEM FAILURE

1234 5*10*2*/ir5 5*10“e/yr

6 5*10"*/yr

7 5*10**/yr

IE RT EP SSR APRS HE FFS

1/уг

* 11 SX

-3J 0.1 1«

10-э

10-Э

10-2

<10

3*10-з

234

567

8 91011 1213 <10:/yr14 З*10*3/уг

15 3*10”*/yr16 3*10,/yr17 3*10’;/yt18 <10"T/ÿr

10"*/yr

10"?/yr 10 I/yr 10”5/yr

10"!/yr 10_J/yr

HE Human Error SSR Secondary Steam Relief FFS Fire Fighting System EP Electirc Power

FIG 7: EVENT TREE FOLLOMING CLASS-IV PONER FAILURE

IE EARNS 1 PP 1 FMP AFNS ECI

123

45

6789

10

10"*/уг

10"*/уг

io:i;/rrЮ J"/уг

5*10"*/уг

5*10-;/уг 5*10 7/ух

РР Pressurising Pumps E&PffS Emergency Active Process Vater System FNP IN/C Pumps HPIS Higt) Pressure Injection System

FIG 8: EV0ÍT TREE FOUMIMG ACTIVE IP PROCESS KATER SYSTEM FAILURE

PS SSR РМР

1

23 2*10'*/yr4 2*10*’/yr

5 2*10~®/уг

6 2*10'П/УГ

7 <2*10*S/yr

8 2*10'4/yr

PS Class IV Povex Supply

FIG 3: EVENT WEE FOUMING NON-АСТПЕ IF PROCESS МАШSYSTEN FAILURE

Cl

APPENDIXSYSTEM RELIABILITIES

52

53

i.iimPDUCTiOH

The objective o£ Nuclear Power Plant (NPP) Safety ia to ensure and demonstrate that the risk from the plant to public and plant personnel is acceptably low. Risk of occurrence of an accident is defined in terms of probability of occurrence of an accident and its consequences in terms of the radioactivity released. The probability of failure in various systems which nay lead to accident situations or affect the sequence of events during accident conditions are evaluated in this appendix.

Although no quantitative risk criterion is specified, a general requirement is that the probability of malfunctions . be limited to small values, decreasing as the severity of consequences increases, so that overall risk remains acceptably low. The following criteria are applied during all stages of design.a. The design, construction and operation of all

components,systems and structures and particularly those essential to the safety of the reactor shall follow the best applicable codes,standards or practice to engineer reliability and safety into the systems,

bi The safety systems shall be independënt of each other and also of the process systems.

c. Each safety system shall be readily testable and shall be tested at a frequency which demonstrates that itsunavailability target is met. As a guide line, a target unavailability of. 10"3 yr/yr is assigned to each safety system. Such reliability goals not only provide targets for equipment design but also for component testing andmaintenance schedules. It is essential that these values be adhered to during the life time of reactor. A deviation could indicate aging of the plant.

2-SYSTEM RELIABILITY METHODOLOGY 8. ASSOMPTIONS

This section summarizes the nethodology of calculation and reliability of process and safety systens whose failure could lead to unsafe situations.

2Л DPLQGI

The computations are based on Fault Tree models. A fault tree is a deductive model which starts with the definition of most undesired event (the system failure) known as the Top Event and proceeds downward till all the combinations of events leading to the top event are identified in terms of failures of basic components. Various stages in the tree are coupled through logic gates.

2b. 2L ASSPHPTIQWS

1) It is generally assumed that the failure rates are constant and the components are utilised during their useful life period only. Equipments are replaced well before the onset of wearout period and early failures are detected during the installation and commissioning phases. The variability in failure rates is treated as lognormal.

2) The number of test operations performed do not cause any significant changes in the failure rates.

3) Piping failures in literature are reported in terms of a)failure per foot-year or b)failure per section-year. Aa any piping system consists of both long piping as well as a good number of sections, an averaging procedure is used in arriving at piping contribution. The total failure rate for per-foot and per-section basis is calculated and geometric mean of the two is assumed to represent the effective piping failure rate.The details of the analyses of the various process and

safety systems of NAPP that were carried out are presented below. A brief system description is also presented herein, however, the details are provided in NAPP safety report volumeI. .

3 PHT SYSTEM PRESSDRE В ш

h failure in PHT envelope would range fron the inlet header rupture to leakage in an inetrunent tubing. A large LOCA is a well studied and defined event and various ESFs are generally designed to cope with this situation. However, various safety studies and operational experience indicate that snail LOCA is highly probable. In addition, a PHWR has large nunber of pressure and feeder pipes which contribute to the probability of failure significantly. In NAPP seperate ESFs have also been providied to cope with the small and nedium sized breaks in the PHT pressure boundary. The major initiating events which lead to a LOCA arei. Rupture.of primary piping including the headers

ii. Rupture of feeder or coolant tubesiii. Opening of Relief valves spuriously or due to a transient

and subsequent failure to closeiv. Rupture of instrument or SG tubing

The response of the various ESFs to LOCA initiating events depends upon both the location and break size and based on these considerations, different LOCAs are categorised as follows:1. Large LOCA > 4я diameter or >10% of double ended inlet header

break area(2A).2. Medium LOCA 1/2 to 4" diameter or 0.1 to 10% of 2A break.3. Small LOCA upto 1/2" diameter or upto 0.1% of 2A break.

3.1 Reliability Analysis

Failures in piping systems are known to have occurred due to a variety of reasons such as design deficiencies, wrong selection of materials, inherent cracks present in the materials, wrong maufacturing, welding or erection procedures, corrosion and mal-opeartion during service. Failures are further grouped into two basic categories -Rupture (catastrophic) and Leakage (non-catastrophic). In a piping which is accessible during operation, non-catastrophic failures would be detected by visual examination or non-destructive testing. In NAPP these have been taken care of by high standards of design and material selection,stringent process control and inspection during manufacturing and installation followed by preservice testing and inservice inspection. Further due to use of ductile

FJQ Ъ'± XRV OPE.N S^FKXISTO OL_OSH.

57materials leakage would most probably precede catastrophic pipe failure. However, this is not always true and some tines a fault/crack may be so oriented as to manifest the failure in a catastrophic manner only. Apart from pipe rupture, a stuck open relief valve will also constitute LOCA. This has been separately anlysed as shown in the fault tree(Figure 3.1).

3.2 Failure Rate Data

The details ol failure rate data are included in [11] which also gives information regarding causes of failure in piping.

3,.3-ReguUB

Using the data as described before,the frequency of failure of various categories of LOCA are as follows:

-4 /yearLarge LOCA »Medium LOCA Small LOCA >The probability of inadvertent opening of relief valve and

failing to close is obtained as 1 x 10“2/year.

« 1 x 10"2/year 1 x 10"2/year

4. REGULATING SYSTEMIn case of any unsafe failure in the regulating system, the

reactor protective system would operate to mitigate an unwarranted situation. Thus the failures included in the failure analysis of this system are those which contribute to a spurious reactor trip. The system is triplicated and eventhough the signal processing part which is based on microprocessor systems is identical to all the channels, channel В is different from both A and C in that it controls the operation of two regulating rods where as both A and C are associated with a single rodeach.

A channel failure results in loss of control on a single regulating rod which can be obviated if the transfer circuit operates and transfers the control to an operating channel.Thus a channel failure occurs when the signal fails and the

transfer circuit also fails to transfer. This is shown In the Fault Tree(Figure 4.1). Failure of two channels is considered to lead to system failure in this analysis.

Ch

<5<hs/*

¡JkXIRGSZSTSH FAXIÆS

ffi/Y’r.

1 1CHARHZLFAXLORS

CHARRO.FAXLORS

CHARHELFAXLORS

F1 ^U J

^До-os/yt

IL

12-5/Гт

Ices/tr

Г—~L

(Л2Кт

<fTIWW ГЖПД1И»

<5satoÍ

<!r ^ ^ trshr 1 2 SXORAL

'JmtVytSERVO SSSVO XSXOHJU.

czscoz* ¿•7/УтД ¿y

£

ff*PROCZBSZRe

^hft

HXOXRORXCBELTA *

«A/Y»

Aff <!> <i> <6 <> 6 <i> <i>CFO DID OXO oac ISO. ARA RISC ООИ

Aie. 0/r CARD

óBTD1

t ;fig, q-iREGULATING SYSTEM FAILURE

:X

mM

en00

59The failure rate data for this systea is obtained fron

[8]. The application К factors in the analysis correspond to an ambient of about 40°C and a general ground based environment and a quality factor value 5 for MIL-e83 grade B1 components. The aean time to repair(MTTR) is assumed to be 24 hours, which is conservative. Also the common cause failures such as power supply fluctuations have been taken into consideration. The failure frequency of the regulating system is 0.3 per year.

5 MODERATOR SYSTEM

Moderator serves as an ultimate heat sink in the 'event of a dual failure involving LOCA and loss of emergency core cooling system. As major fuel failures can be averted for extended periods due to presence of moderator(upto full level in calandria)[4], it is essential to ensure that this systea is reliable. It would be seen that a) the frequency of a leak in the system that affects the moderator level is low and b) in case such an incident occurs it is possible to isolate the moderator system so as to stop the leak.

5.-1 SYtttcm DescriptionModerator system consists of two independent loops

connected to the calandria vessel. Each loop consists of a set of two pumps, a heat exchanger and associated piping and valves. There is a standby pump which can be started when any of the operating pumps in any loop fails. For the full power operation of the reactor it is essential to ensure that both the loops are operating. This system also caters to absorber; regulating and shim rods. Required quality of the moderator is maintained by the purification system, which takes continuously moderator from the system, purifies it and returns it back to the main moderator circuit. To keep the required reactivity level of the system, poison concentration in the moderator is controlled by the poison addition system and purification system. Moderator purity is monitored by the sampling station.

3.2 Reliability Analvalw

As the system is a continuously operating system, the frequency of moderator system failure leading to a decrease in level is a measure of the reliability of the system. As maintaining moderator level in calandria has an impact on fuel safety in accident situations, the probability of failure to isolate the moderator also serves as another index of reliability. In this analysis these two are considered.

S.a.ABgUBPtlongThe analysis is carried out under the following

assumptions.1. For heat exchanger tubes only rupture is considered to lead

to moderator level drop. It is assumed that leakages do not affect the moderator level much.

2. For large diameter piping, leakages in addition to .rupture are also considered as these can result in significant losses from the system. To take care of this the upper bound failure rate of rupture is used.

3. For the bellow seal valves only failure modes resulting in external leakage are considered.

4. A single pump seal leakage, it was observed in reference [12], results only in a loss of «15 litres per minute. Even if all the pump seals leak simultaneously, the total leak rate is «65 litres per minute. Based on a rough estimate[12], it is found that there is about one hour time available for moderator level to reach the calandria isolation limit in the event of all five pump seals simultaneously leaking. Operator action, to isolate the leaking pumps, is highily probable during this one hour. Hence pump seal leakages in moderator system are not considered to lead to a significant level drop in the calandria. The length of small and large diameter piping considered here is shown in table 5.1.

5. For diaphragm valves in moderator and associated systems failure rate data available from MAPS operation[13] have been used, details of which are shown in table 5.2. As failure information on bellow seal valves is not available, about one third of diaphragm valve failure rate is assumed to be

61applicable to these valves.

The failure data used in this analysis is shown in table5.3. A fault tree for the moderator level drop and failure to isolate moderator in calandria have been constructed and shown in figures 5.1 and 5.2.

5.4 Résulta

With the above assumptions, using the data shown in table5.3, the frequency of moderator leve.l drop has been calculated and this turns out to be 2.0/year. Th« probability of failure to isolate the moderator in calandria turns out to be 5.0*10‘3.

Table 5.1 Details of Piping of Moderator ayates

System Lineno

Length in Metres

Number of Sections

Mpdmtpi .CirculationSesteaa)Piping & Tubing All lines 152.7 59

(diameter >3") b)Piping & Tubing 3211.8 14.0 18

(diameter<3a) 3211.9&Ins >Shim Regulating RodCooling System

Tubing

All 3240 42.5 13(diameter >3")

Liquid- PpjgpD-.Sya tealines3481-1,2,1 76.5 15

(diameter<3a) Semolina circuit

12,13 & 1610.0 1

(diameterO* )Bmificatlpn Circuit;

(diameter <3a)3221-1&2 (100\ length) Remaining all 3221 lines 50% length (The port­ion beyond MVs uptoIX is not considered

•

Table 5.2:Diaphragm Valve failure data MAPS 1 Operation

Description Value

Number of Valves 486

Number of Diaphragm Replacements

83

Percentage of failures assumed to lead to major External leak

50%;

Period of operation(years) 3.25

Failure Rate(per year) 0.03

Table 5.3; Failure Rate Data

Item(Failure Mode) Failure RateElPC/Xufce

{Rupture/1eakage) i) <3"diameter

ii)>3"diameter

HfiAlk Exchanger(Tube rupture)

£ШШ (Seal leakage)

Manual Valvea(External leakage)

MQtOI Operated Valves i) External leakage

ii) Failure to operate

&ÍX Operated Valves i) External leakage

ii) Failure to operate

Check ValvesExternal leakage

Control ValvesExternal leakage

Ilansis. joints

3.0*10"*/section-year 2.0*10^*/foot-year 3.0*10"*/section-year 2.0*10“3/foot-year

8.8* 10'® /tube-year'

0.108/year 3.0*10"2/year

3.3*103?/year 1.0*10’ /demand

6.0*103?/year 1.0*10" /demand

2.6*10"3/year

6.0*10’’/year 1.0*10"2/year

/;

V39 MV28 V38 CBS>

V37 V42 V34 . »43( BS> <BS)

»46<B8>

»17 MV2.1fV4. . . . ,MV16

M»1.HV3. . . . .H»13

HV27 »29 MV33 t BS> ( BS> < BS> в

FIS. 61 FAULT TREE FOR 'DROP IN MODERATOR LEVEL

C3

IO

ra

ИОВЕНАЗОВ CIRCaUkXZO« STSTZH

0-3/Yf à

i i —1Í VALVES 1

:jPOMP BEAT

! t '

ù

О О О О О ÖCV60 CV61 V40 V39 va V9

(BS> <BS) (BS) <BS>

ЛI

Ä

; рш» i ¡ 1 ! ‘ POMP

? г ;I POUF j POMP VALVES

3-e 1 4 ¡ 9£ ^ A A A Ai

vaoо Ó 6 ó ó ó óV4 V5 MV1S HV16 V24 V25

<BS> (BS>

FIG. ( A) FAULT TREE FOR " DROP IN MODERATOR LEVEL“

)

■ LIQUID POZSOii II STSTSM I

0-27/tf. 7b\

HV26 VI V38ÓV310 Ó Ó Ó ÓV24 V36 V37 V39 V2

!Ovs»

SHZM JUm RBCTUI.ATIWG BOD COOLIBO SYSTEM

09S/Yt 7ü\

' ABSDEBER I ! ABSORBER ¡ ! ABSORBER 1 j ABSORBER! ROD Al ROD A2 ROD A3 ROD A4

AЛ\I

л\

■ HAIN SYSTEM Ij Valves j

; REGULATING ; ! ROD Rl !

REGULATINGj ROD R2

SHIM ROD SI

Л ' i . Л¡Aл\ A

ÔÔÔÔÔÔÔÔ ¡ i i : ¡A. .^4

vil V12 V13 V14 V23 V29( as> <bs> <33> (bs> <bs>

и и и и и WV37 VI V2 V7 V14 V9 VIO V13

< BS) < BS) < BS) C BS> (BS> CBS) ( BS)

SHIM ROD 52

лЛ».4_i

HG. S-l í В, C ) FAULT TREE FOR “ DROP IN MODERATOR LEVEL“

09

. CALENDRIA ISOLATION FAILURE

5 >10if

-a

Ó Ó 1о3211 3211 3211 3211MV2 MV3 CV60 CV61

FIG. 5.2 FAULT TREE FOR " CALANDRIA.'lSOLATION FAILURE"

09

69

6 FEED WATER SYSTEM

Feed water is essential for heat renoval fron Prinary Beat Transport System(PHTS) when the reactor is operating or shutdown. Beat removal fron PATS is acconplished by nain feed water system(MFWS) when the reactor is operating while auxiliary feed water system caters to conditions of MFWS nonavailability and during reactor shutdown. The nonavailability of MFWS results in high PBT pressure and consequent reactor trip where AFWS nonavailability results in loss of heat renoval capábility fron FHTS and thus has safety inplication. In this section an evaluation of the frequency of MFWS failure and the demand failure probability of AFWS are presented.

^I Svsten Description

Feed water system consists of three 50\ main boiler feed pumps which take suction fron the deaerator storage tank and supply water to the four stean generators after passing through & drain cooler and a set of heaters. The flow to each of the SGs is regulated through a small and a large control valve each of which has a standby to take care of any maintenance of the operating valve. Only two main boiler feed punps are enough to supply the required feed to the SGs: However, when any of the operating pumps trips, the standby comes automatically. Even when the standby does not start, operation of the reactor at reduced power is possible and this requires operator action. This operation at reduced power is not considered as it demands operator attention and alertness within the first few seconds of the standby pump failure when any of the operating punps trips. The level in the deaerator tank is naintained by a nain condensate extraction pump(CEP) which has a backup. The condensate pump discharge is preconditioned to deaerator water temperature through three LP heaters. One of the two AFWS punps starts automatically when the main boiler feed punps trip and supply water to the SGs for heat removal. Both MFWS and AFWS share some common piping. In addition AFWS has an independent line which feeds water directly to the SGs.

fi.2 Reliability Analvala

In carrying out the reliability analysis the following assumptions were made1. Only one of the two HP heaters is sufficient for normal

reactor operation.2.Two out of the three LP heaters are enough for normal

reactor operation.3. Auxiliary condensate extraction pumps are not considered

here as these do not have any safety implication. This is due to the adequate capacity of the deaerator storage tanlt which is fed by these pumps.

4. Valves in feed control station associated with each of the SGs are not considered as closure of these valves affects only one SG and this might lead to a reduction in power. Simultaneous closure of the valves in all the four feed control stations is remote. However, these are susceptible to common cause failure due to common power supply.

5. Only major leaks from flange joints(mainly from condensate éystem) are considered to affect the system performace.Based on the above assumption, fault trees for the main

feed water system failure and auxiliary feed water system unavailability have been drawn and are shown in the figures 6.1 and 6.2 respectively. The details of 'the piping are considered in this analysis are shown in table 6.1. The failure data used in this analysis are presented in table 6.2.

The frequency of MFWS failure and the demand failure probability of AFWS have been calculated using the fault trees(fig.6.1 and 6.2) and using the data shown in table 6.2 and these trun out to be 0.5/year and 1.0*10"3/D respectively.

71TABLE 6.1: DETAILS ОТ PIPING OF FEED WATER SYSTEMS

SYSTEM Length In meters(s ixe)

Number of sections

Number of flange Joints

Condensât» System 201.64 33 ,37(die > 3")

Feed System1. Independent 131.3 37 -

piping (día > 3">

to heater no.6)11 » Common piping 169.6 62 -

(dis > 3")Steam piping 308.8 12 -

(día > 3“)

¡/

72

TABLE А.2* FalLurff Rat» Data

ltam(Fa1Lure moda) Failure rataPumpa(Including motor)

1. Falls to Start 1.0*10"3/D11. Falls to run 3.0*10"5/hr

Haatar1. Shall Leak 1.0MCTe/hr

11. Tube rupture (1000 tubes) 0.01/year

Motorised Valve1. Falls to remain

open 1.0*10"4/D11. Failure to open 1.0*10_3/D

111. Failure to operate 3.S*10"3/DManual Valve

1. Falls to remainopen 1.0*10"4/D

11. Failure to close 0.036*10"®/hr 2.7*10"'/hr111. Stuck closed

Level Switch1. Failure to operate 1.0*10"4/D

<3.0*10"‘/hr)1.0*10"4/D<3.0*10"7/hr>

Pressure Switch1. Failure to operate

Hand Switch1. Falls to transfer 1.0*10"5/D

Limit Switch1. Failure to operate 3.0*10"4/D

Relay1. Falls to energise 1.0*10’4/D

11. Failure of NC by opening given no 3.0*10"®/hr

<1.0*10"5/D>switch operation111. Failure of NO

contact to close when energised 3.0*10"7/hr

1v. Failure of NCcontact by opening or coll openingor short 1.0*10 7/hr

TABLE 6«2i Continuad

ItamtFalLura moda) Failure rataPiping ruptura

1. (día < 3"> 3.0*10 уsec-year2.0*10"J/ft-year

1. (día > З") 3.0*10“*/sec-year2.0*10"*/ft-year

Strainer1. Plugged 1.0*10“*/hr

Expanalon joint 4.0*10“3/DControl valva 1.0*10'3/DFlow element

1. All modes 0.243*10 e/hrHuman error 1.0*10"3/DFlange joints 1.0*10“‘/hrCheck valve 3.0*10“’/hrCircuit breaker 1.0*10“3/D

1.0*10“®/hr

j LOSS OF MAIN FEED WATERA0'5/Yr

I • ~=Tt

l

соыФЕ.ы%>чте.* •

SvSTE. FAXUOtlE. PUMPS j i PIPINGI I

jHEATERSA ? ? I

! : ! ! i .t

Л/521

ASSE

DESIGNEDOBS

DESIGNED

«,-4r

PUMP RECIRC. PIPING

FIG. 6.1 LOSS OF MAIN FEED WATER

ЛСТОДХХОЯ-]LOGIC !

'

/'S . . V J

PGMF OPERATING FAILURE

POMP DEMAND FAZLORE

ьо Ó' О-

•ОSTR1 EJ3 VI V22 vai

<SUCTION) <DISCHARGE) CHECE VALVE

FIG. 6. la LOSS OF CONDENSATE SYSTEM

XX.

oí

6 TUBE

ЯВРТОВЕ

4

Ô

ISOLAT! OB FAXLOHS

ЛI

HEATERS

ó - óV33 FAILS OPERATOR V32 FAILS TO CLOSE ERROR TO OPEN

Aо

1f ЛiLP 'HEATER 1 ! LP

HEATER 3. LP i! HEATER 2 .?

▼26 FAILS TO CLOSE

r\V35 FAILS TO OPEN

ÔÓ

й

О

V34 FAILS TO CLOSE

ÓV36 FAILS TO CLOSE

Г-'îLS- 80

FIG. 6.1a (Ci) LOSS OF CONDENSATE HEATERS

i

OPERATORERROR

PI

TRIPPED POMP 1¿s

STOCK CLOSED

к\^,t HS1076

HUMAS ERROR

f ?Г-'svees

—\Ô

svaee

FIG. 6.1o (Cz Cs) FAILURE OF CONDENSATE SYSTEM

■M•M

POMPS ¡2 хю-^Г

} d)CIO~A/VvPOMP

FA1L08E I Л "! Ixio'VyV

àCLASS ZV

POSER SUPPLY

НА1МТЕЫАНСЗ

TTi.¿xiO’fc/d

P6 FAILURE TO START

“■O-58/Гг I

OPERATING POMP FAZLORE

A A

\ПюгтACTUATION О

POMP

0-27/Yi

A

PSMAZNTENANCE

Â

PS! FAZLOREI

<л/гЬ

?4MAZNTENANCE

00-23/Гг i 1

j P4 ” i ! ps/61 FAZLORE FAZLORE ! FAZLOREA ■

!

ñ|>

- 4 ?4MAZNTENANCE

*6 iFAZLORE Í

A/"N

PS LSÓ

OVERLOADRELAY

FIG. б. I (FI) LOSS OF MAIN FEED WATER PUMPS ■MЭЭ

4..*

¿•8iuö4Av^—A

5 ïMAINTESASCS SPURIOUS

HSA3SB Hsaxss HEATERб HEATER63

5

VALVES

XHEATER

S

A A A ¿T*<ToMV MV 164 166

6ISOLATIOMFAZLOREz

6 Ъ 6 ъHEATER 6 HEATER 6 OPERATOR LS PREMATURE DORM FAILS ERROR CLOSURE

ó ó ' ó ó Ъ

OPERATOR LS MV-163 FAILSFAULTS TO OPES

MV-166 PAILS TO CLOSE

MV—167 FAILS TO CLOSE

FIG. 6.1 <F21L0SS OF MAIN FEED WATER HEATERSсэ

rOsS“OsS

VALVE OPERATOR LS PROTECTION MV FAILS В PASS FAILSFAILS ERROR FAULT ON P-б TO OPEN TO OPEN.

FIG. -6.1 (F5) FAILURE OF MAIN FEED WATER PUMPS.

oo

LOSÉ OF AUXILIES!feed gaaot

! 1х|о"3/Уi '■

FIG. 62 LOSS OF AUXILIARY FEED WATER

00

i VALVES

2?COMMON PIPING

VALVES • INDIVIDUAL PIPING

CV229 STOCK CLOSED

ó4

VALVES

VI06 HUMAN EKROR.

гФ<IO-*/4

V107 HUMAN ERROR

VI12 HUMAN ERROR

MV-183

ЛMV-184

Ô ÓCHECKVALVE

MV-183

FIG. 62 (Аг) LOSS OF AUXILIARY FEED WATER

' - .s.$

• i

•I

09 .

ACTOATOR

ил

Ш)сР/4 Лсл

л/А!

HOMAN ERROR SWITCH

6 ó Ъ IОP4 СВ

CONTACTP5 CB

CONTACTP6 CB

CONTACTVALVE FAILS TO REMAIN OPEN

CB CONTACTSo ö 74

V104

HS746 LOP

13-4x10-3/4

Д

оLIMITSWITCH

óHUMANERROR

FIG.. 6.2 (Аз) LOSS OF AUXILIARY FEED WATER

CoCO

7- PROCESS WATER SYSTEM

Process water system liAce electrical power supply system, is & support system. It does not directly affect the safety of the reactor but it affects the performance of the safety systems. Hence it is essential that this system be highly reliable to prevent the dependent failures of the process/safety systems. In this respect the failure frequency of the process water systems (which lead to reactor shutdown and has a bearing on the ultimate heat sink of the reactor) and the demand failure probabilities of emergency process water systems (which affect the safety systems) are analysed.

7.1 System Description

Process Vater Systems are designed to remove heat from various process systems, like moderator system( through moderator heat exchangers), shutdown cooling systems(' through shutdown coolers), safety systems like ECCS( through ECCS heat exchangers) and service systems like class III power( DC jacket cooling). To cater to the various active and nonactive systems, the process water systems are classified as:1. Active High Pressure Process Water System(AHPPWS)2. Active Low Pressure Process Water System(ALPPWS)3. Non-active High Pressure Process Water System(NAHPPWS)4. Non-active Low Pressure Process Water System(NALPPWS)

Out of these the NALPPWS being a non-safety related system( in the sense that it does not cater to any safety system) is not considered here. AHPPWS and ALPPWS systems are not normally active. But as these systems cool active systems like PHT pump gland cooling, moderator cooling, these are likely to contain some activity in case of leaks from the active systems. These active systems, are further cooled by active process water cooling system(APWCS) which mixes with the NAHPPWS before going to induced draft cooling tower(IDCT) for heat removal.

85^The flow requirements of these systems during normal

reactor operation and shutdown operation are different except for NAHPPW system. In case of ALPPWS and APVCS shutdown requirement is less than normal and it is met by seperate emergency pumps primed by class ITT power. In case of AHPPVS, shutdown requirement is more than normal requirement and the same type of pumps with a more number of pumps running! on class III) meet the emergency requirement. In case of NAHPPWS normal and shutdown requirements are the same and these are met. by the same type and number of pumps.

1.2 Reliability AnalvaljThe fllowing assumptions are made in carrying the analysis.

1. In Active Process Water Cooling System there is one standby pump for both the units. It is assumed that the demand on the standby does not arise simultaneously for both the units.

2. In Emergency Active Process Water Cooling System it is assumed that the number of pumps required are always four.(When the standby cooling system operates the pumps required are four)The piping data giving the piping details inside and

outside RB for large( > Э* dia) and small ( <3" dia) piping are given in table 7.1. The failure data used for this analysis is shown in table 7.2.

7.3 ResultaFault trees (figures 7.1 to 7.4),.' as it was mentioned

earlier, were drawn for Dthe frequency of process water system failure and 2) the probability of failure on demand of 2a) emergency active process water system and 2b) emergency nonactive cooling systems. Using the fault trees and utilising the piping data in table 7.1 and failure data in table 7.2 calculations have been done and these turn out to be 2.5/year ,4.5*10"*/D and 1.0*10"4/D for 1,2a),and 2b) respectively. The contributions of piping to the failure frequency are shown in table 7.3.

TABLE 7.1: PIPING DETAILS

Inaide Reactor Building Outside Reactor Building

Spatem Length No. of Ro.of Length No. of No. ofin Sectiona Flange in Sections Flangemeterá jointe meterá joints

AHPPWSdie < 3- 377.0 61 50 - - -die > 3- 221.0 4.8 46 103.0 10 8ALPPWSdia 4 3a - - - - - -dia > 3B 43.0 6 - 2748.0 176 153

RAHPPVS -dia < 3a 13.5 4 60.0 2 -dia > 3- 630.0 - 42 38 1010.0 37 24

APCWSdia < 3* - - - 67.0 22 22dia > 3* — — 515.0 75 50

TABLE 7.2 : FAILURE RATE DATA

Iten(Failure node) Failure rate

Punps(Including notor)i. Fails to Start 1.0*10'3/D

ii. Fails to run 3.0*10”3/hrHeat Exchanger

8.0*10’®/tube-yrii. Tube ruptureMotorised Valve

i. Fails to renainopen i.omo’Vd

ii. Internal leakage 3.0*10’®/hr(Catastrophic)

iii. Failure to operate 3.5*10”3/DManual Valve

i. Fails to renainopen 1.0*10”4/D

ii. Failure to close 0.056*10”*/hrCircuit breaker

1.0*10’®/hri. Spurious transferii. Failure to transfer 1.0*10"3/DStrainer

i. Plugged 1.0*10"S/hr4.0*10’3/D

Hunan error 3.0*10"3/DExpansion joint 4.0*10"3/D-

1.0*10"5/hrFlange joints 0.3*10"2/yr

TABLE 7.2: Continued

Iten(Failure node) Failure rate

Piping rupturei. (dia < 3") 3.0*10"4/sec-year

2.0M0"4 /ft-yeari. (dia > 3a) 3.0*10"5/sec-year

Check valve2.0*10"5/ft-year

i. Internal leakage (Catastrophic)

ii. Failure to senain

3.0*10"T/hr

openPressure Switch

1.0*10"4/D

i. Failure to operate

Linit Switch

i.o*io"4/d(3.0*10"’/hr)

i. Failure to operate Tank

3.0*10"4/0

i. Rupture 1.5*10"9/hrClass III Power Supply 5.8*10"3/yrClass IV Power Supply 1.0 to 2.0/yr

TABLE 7.3:PIPING FAILURE FREQUENCY

Systea Inside RB Outside RB

Active Process Water systemDia <3- 0.067/year

Uia >3- 0.16/year 0.56/year(Inclusive (Inclusiveof joints) of joihts)

Active Process Water Cooling SystemDia <3- 1.7*10"2/yearDia >3* - 8.75*10"3/yearNon-active HighPressure ProcessWater SystemDia <3- 3.3* 10°/year 4.9*10"3/yearDia >3" 7.1*10'3/year 9.0*10"3/year

АРИ SYSTEMГ«/*

I3xi0's/YrTANKS !

I 13/Yv

4>

COMMONPART

XT

АНРРИSYSTEM

ЧО-^/Yr 7X10"^ /fr.

1

~ Xù'Wrr jO-7/YrACTUATION .< \*. »• HXs

LOGIC PIPINGX

\x-CLASS ZV

POSER SUPPLYUli

Xл

/г'ДЗ

SJ1202

X,w'

EJ 1209Ó

V1213Лч_

V1202

XНК

TOBE•fr FOR. THE STANDBY HEAT EXCHANGER

\/ALVES HAVE TO EE OPENED AND HENCE TKE ÎÆMAND FAVLURE P?DEAft\HTV FOR VAM-VES ENTER .

Г i

ALPPSSYSTEM

/Д2

1 1 X 1 1Í7131Í ! HX1 I 7131

HX27131HX3

7131HX4

7131HXS

7131HX6

7131HX7

/А1

COо

FIG. ACTIVE PROCESS WATER SYSTEM FAILURE.

SIGNAL

i POMPSj PIPING &

VALVES

i 17133 jPI ! 7133

P27133P3

7133P4

7133P5

А л^Т А' А

0+ Ö Ô!1

О (ÿV1202 V1004 STR POMP CB CHECK

1202 VALVE

we lNCLU*Deb IN COMMON PIPING t FOR. STANt>b7 PUMP ALSO

PAVE TO CHANGE STATE.

FIG, T-.HAV ACTIVE HIGH PRESSURE PROCESS WATER SYSTEM FAILURE. CO

EJlOO1 EJ 1215 STR1201

V1208 MV1021 + HAND SWITCH

POHP CB HV1021LS

77V120Ô

LS

* THESE HAVE TO CHANGE STATS FOR- STAN'Db'V PUMP OPERATION AND A U SO CHANGE STATE \WHEN THE OPERATING PUMP TRIPS .

НОМАМERROR

FIG. 7.KA2) FAILURE OF ALPPW SYSTEM

БДРВSYSTEM

LEAKAGE OPERATORERROR

EJ1202

EJ1207

V1213 V1202

FIG. ?,2 FAILURE OF EMERGENCY ACTIVE PROCESS WATER SYSTEM.

ЕАНРРвSYSTEMw*

с \**-

PIPING & VALVES

lixio

2POMPS ACTUATION

LOGIC2-2X10

•1----------- 1 1 ___u,

7133PI

7133P2

7133P3

7133P4

7133PS

¿__iA

A

XwPOMP

X■ i

SIGNAL

ъ

V1202 V1O04ОSTR1202

V1003 CH. V.

Ъ

CB

"ÔHUMANERROR

FIG. V--2 (EA1) FAILURE OF EAHPPW SYSTEM

V1002 VI211

(УVALVES

ЕА .LPPff SYSTEM

< IXio-4/У

5

7131 P3

74s

Ъ

HS HUMAN ERROR

ÔCBs

ЛPUMP

7131 P4

Ï&

àSTR1203

1*10PUMPS ACTUATION

& LOGIC

1f \KJ

V1001 CH. V.

FIGURE- 7.21EA2) FAILURE OF EALPPW SYSTEM.

оPOMPOP.

1 ~6xlQ-6/d

I/StuL__i

Л

EMERGENCY NAHPPW COOLING SYSTEM

IXIO■-+U

~ 4*\о~ь/а

ФI ЭЯ2

/ЕЯ2

О oPUMP DISCHARGE ACTUATION

DEMAND VALVE MVIOOI LOGIC

ЛCB

] i-6xiEAPWC NAHPFWS COMMONSYSTEM PUMPS PART

X

7134 PI 7134 P2 ACTUATION IDCTLOGIC

( ОOPERATOR ERROR TO

ACTIVATE PRELUBRICATION

* FOR n°n operating pump

o~s/d

POWERSUPPLY

FIGURE- 1-3 FAILURE OF E NAHPPWC SYSTEM

PI

PUMPPONCTION

FIGURE-

\d44

POMPDEMAND

ACTUATIONLOGIC

DISCHARGEVALVE

?.3(EN1) FAILURE OF E APWC SYSTEM

-з,I«io Ht

HOa-tCTZVZ COOLING SYSTEM

ГО-3/Yr.ЛРЯС

SYSTEM

JïMftrNAH?FW SYSTEM

Л А т

I 1 i 1HE'S О V

PIPING CLASS ZVPOMPS

POWER SUPPLY Ф

NALFPWSYSTEM

PIPING

6 ■ 6 Ó Ó Ó Ó ÓШС1 ISOLATING HE2 HZ3 ИМ HZS HEB HS7

2 VALVES -DO- -DO- -DO- -DO- -DO- -DO-

P1 P»

Л6 5 ó ó ó ó ó

SP.XSZPSIGNAL

DISC».VALVE

CB LFO POMP STB DISCH. VALVE • OP. FAILS TO CLOSE

óÔ

ACSiam.

àFOHP

DEMAND

óPDMPOP.

Ó 5 óDISCH.VALVE

DISCH.VALVE

STBAINSB

COOMOHPABX

Л

ACTOATIOHLOGIC

ъIDCT

* Ifc \S OSb«jm«d demand О»» 1*»« gbxindW ^UYn^. "iS Ykct SÁ-rriu¿a**<-0 ч-С

jar boTW 1h€ Unib .

3LOP СЭ

F16. 7-4 FAILURE OF NONACTIVE СООШС SYSTEM СО09

НАНРРЯ

0.2иО.«4ЖР—TÍci

О ‘-*<55-2 Xlûs/yf.

CLASS ZIZ POOER SUPPLY

5CBot.

3-6>lO~4 /y-f.PDHPS

■IbùtA-hr.

PIPZNG

TU

PUMPOP.

ACTUATIONLOGZC

P-l

"5 JMVlOOl CB

DEMAND

"ff

POMPDEMAND

^ APPL\CAB ^ SfANt> Рим P ONL>/ .

OPERATOR ERROR TO ACTZVATE PRELUBRICATION

FIGURE- 9.Í.IWFAILURE OF NAHPPW SYSTEM

100

fl-COMPRESSED AIR SYSTEM

The Compressed Air System comprises of six compressors and three dryers for both the units along with the associated air receivers,valves and piping. During normal reactor operation,three compressors are connected to one common header and the other three to another common header. The headers in turn are associated with a dryer each and in case of failure or maintenance of any operating dryer, the third dryer, can be valved in. Two compressors in each unit are 'ON1 during normal operation, one being on class TV and the other on class ITT and the third compressor is rtandby on class III. It is usually a unitised operation with tie up valves V-1030 and V-1029 kept normally closed.

a.J-BgHafcültY„&nalygia

The loss of compressed air situation arises when the air pressure falls below 7kg/cm2(g) in the common header. The details of reliability analysis are shown in the fault tree(Figure 8.1). The failure modes for loss of instrument air are as follows:a) Two out of three compressors trip or one of the running compressors fails and the standby compressor fails to start on demand or fails to run. The running compressor would be able to maintain the air pressure above the limit for about 10 mts. The tie line valves must be opened within the duration manually.b) Failure of one of the dryers coupled with the standby dryer

either under maintenance or failure 'on demand* or failure in operation during the maintenance period of the first dryer.c) Failure in piping.

8,2 fflUme Rflte-Pflfca

The Air compressors may fail due to the failure in any of the following components.

101i. Suction filter

ii. Compressor or the Motoriii. Inter cooleriv. After cooler

V. Inter connecting piping, tubing etc. vi. Air receivers including the relief valve instrumentation

tubing, piping etc.In case of the Air drying plant, various components leading

to dryer failure are:i. Regeneration system components,eg RV-1726,V-1689/V-1688 or

V-1692/V-1693, blowers etc.ii. Prefilters and associated valves,postfilters and

asssociàted valves,iii. Piping.However,the reliability analysis is based on the failure data obtained from RAPS [14] wherein the availability figures for LP and HP compressors and dryers are included as total subsystems. The fault tree is also not developed down to the component level due to the same reason. Based on RAPS data, following failure data has been used in the analysis.

Compressors: Availability - 90\Failure rate - 3/yrMaint, down time - 15 days/yr

Dryers: Availability - 90\Failure rate - 2/yrMaintenance down time - 15 days/yr

Class III Emergency Operation of Compressed Air System( Instrument Air)

During class IV power supply failure loads are automatically brought down and handled by class III power supply,wherein only one compressor is 'ON' (one which was working on class IV switches over to class III,ie CP2) and CP3 will remain as standby. Similarly air drying plant DR1 will switch over to class III and there will not be any standby drying plant during class III operation assuming unitised mode. These conditions have been explicitly shown in fault tree (Figure 8.2).

FIG. 8.1

COMPRESSED AIR SYSTEM FAILURE(INSTRUMENT AIR CL IV)

>=* 2/YyU »7«IO'2

AIRCOMPRESSORS

1 4-5/Yr. i

? 10 ORSER

RUNNING

ъ

STANDB7

1 0-5OPERATION О MAINTENANCE

PIPING COMPRESSOR

Д

¿1Л

¿if"; ó 6 ó

CflFAILS

CP2FAILS

СРЭ FAILS ON DEMAND

CP3FAILS

RUNNING Ó

1 1CPI CP2 CP3

CP2 UNDER MAINTENANCE

FIG. 8.1 (A) COMPRESSED AIR SYSTEM FAILURE оGO

FAILS ON FAILSDEMAND

DR1 DK2 UNDERFAILS MAINTENANCE

FIG. .8-1 (B) COMPRESSED AIR SYSTEM FAILURE

COrWÍESSCD air SrSTCT F*IL'JQ£ lINSTeur.£W AIR OASS Ш)

( CLASS IV VRAILURg Г

”77 777 10

1.0 X 10'-3 VO0-’ óDA 1 PYPTtir

г.о ж 1cr2^Lwr* * » Д- *“3

FAILURE COMPRESSORS CLASS III FAILURE

5CP 2 FAILURE

CP 3

? ó ó óСРЭ CP 3 CP 3

FAILS 04 OERANO FAILS IPtOER I1AINTENANC£ rls« 8-? COHPflESSEO AIR SYSTP1 FAILUAC ( IICTRUPIEMT AIR CLASS III )

L.3 RcBUlta106

The frequency of compressed air failure in the unitised mode of operation is 2*6/yr and the probability of failure on demand of class III instrument air system is 1.0*10‘4/D.

Д-ELECTRICAL,POWER SUPPLY SYSTEM

The electrical supply system which provides power to all station loads, is an important safety system since it is essential for the satisfactory operation of various other safety related systems. The system is broadly classified into ‘four different categories of power supplies depending upon the reliability, continuity and availability of the power supply requirements. These are termed as -Class I, Class II, Class III and Class IV supplies. Class IV and Class III are the basic normal and emergency power sources respectively for long term operation.

9.1_, Class IV Power Supply

This forms the main source of power to all the station electrical loads under normal operating conditions of the unit. There are two diverse and independent sources of Class IV power, one from the 220KV grid through a 220/6.9KV start-up transformer and other from the station generator through a 16.5/6.9XV unit transformer. The two sources are interconnected(at 6.6KV level)' in such a way that in case of loss of power from any of them, power supply can be maintained by a fast automatic transfer of loads to the other healthy source.

9.1.1__Reliability. Anataaia

The failure of Class IV supply is not unsafe, however, it is an important initiating event since a number of process systems connected on Class IV supply e.g. PHT pumps. Main BFP etc. trip and DCs are the main source of supply to emergency loads on Class III till it is restored.

The frequency of Class IV supply failure is the iaportant parameter which, coupled with the unavailability of Class lit supply, would yield the frequency of Station Blackout after about 30 mts. of loss of both the supplies. The reliability analysis of Class IV is shown in the fault tree of figure 9.1. The dominant failure modes are*.1.Failure of grid supply when station generator or unit transformer is down due to maintenance

2. Failure of station supply when components of grid supply are down

3. Simultaneous failure of both the sources of Class IV supplyThe frequency of Class IV failure has been worked out using

Markov Techniques[9]. It is important to note that it would be essential to consider both the effects of grid fluctuations on the performance of reactor and vice versa - any transient leading to reactor trip and subsequently, disturbing the grid stability. Both the situations would lead to a Class IV failure.

9^2 -_CLASS III POWER SUPPLY SYSTEM

During normal operation, Class III buses are supplied from6.6 KV class IV buses. When Class IV is not available, the loads on Class IV are automatically dropped to enable the DCs to start and subsequently, pick up in sequence after DCs are at rated speed. The transfer is affected through the Emergency Transfer System. The sequence of pick up is chosen with reference to the urgency and importance of each load. Normally more load is connected to Class III buses than that could be handled by a single DG. This is permissible as the DG rating is higher than ¿he nominal during the first two hours of operation. In the event a single DG only starts the sequence of load picking is stopped at a condition where the operating DG is not overloaded. The system is considered safe even if only one DG is in operation and the priority loads are connected. The tripping of isolating CBs and the closing of CB in series with the DG is done by interlock circuits and their availability have been considered in the analysis.

3^1 RflllftbllltY Analvala108

The details of reliability analysis are shown in the Fault Tree of figure 9.2. The major failure modes of Class III unavailability area. Independent failures

Since the emergency power supply system comprises of 2(100\) DCs, availability of any one of them would, as mentioned before, be adequate. Thus, failure of both DCs on demand would be the failure criterion. The contribution from the bus and the CB associated with each DG is also included. The operation of the CB is through interlock circuits whose contribution is associated with the CB.b. Test and Maintenance

Testing and Maintenance of components contribute to the system unavailability due to reduced redundancy during the test or maintenance period and is also a function of corresponding intervals. Test contribution would, however, be negligible, к downtime of 7 days/year has been assumed for a DG in the calculations. Thus, the maintenance contribution ■ 2*7*(Qd+Q0)/365 where Qd is the probability of faiure on demand and Qq is the probability of failure in operation during the mission time( assumed to be 24 hours) when the other DG is under maintenance.

9.3 Common Cause Failures

Common Cause Failures(CCFs) are multiple failures which are dependent and caused by a single initiating cause. Various factors contributing to CCFs in DCs may be listed as follows:a. Design and Fabrication deficiencies e.g. fuel oil blockage,

water in fuel oil, common service water supplies and D.C. supplies,

b. Operator errors in test and maintenance,c. External Environmental Effects e.g. rise in room ambient

temperature.

109In case physical diversity and fire barriers are provided, the effects of CCFs emanating from the external environment e.g. fire, change in room ambient temperature etc. would be reduced. Since the maintenance of two DCs is independent and staggered, the contribution due to human error in test and maintenance is significantly reduced. It has been bbserved[10] that lach of detailed procedures,checking the restorability after test and maintenance are dominant causes of CCF due to human error in test and maintenance. Other factors contributing to CCFs area.fuel oil blockage or water in the fuel oil system,

b. lack of water chemistry control in the engine jacket water causing corrosion

c. service water system or DC power unavailability,d. loss of start air pressure etc.

The overall contribution due to CCFs is quite dominating and is estimated as 1,0*10°

Э.l-Overloading EffectsOn Class IV failure, all the loads connected to Class III

buses are dropped. When DCs start and pick up speed, the loads are sequentially picked up in accordance with the emergency transfer logic. In case a CB associated with any Class III load fails to trip, the corresponding load would not be dropped and the DC may be overloaded and because of the intertie, the other DG could also trip. But the DGs are designed for a minimum of 10V overload and no single load exceeds this capacity. Thus, at least two CBs must fail to open to cause any overloading of DG. The probability of this failure is I.OMO*4.

9.5. ClftBS—U .SuREly

Class II is the uninterrupted supply required for the important systems like reactor protective,regulating systems etc. The Class II buses are normally fed from Class III through ACVRs and the MG sets and during nonavailability of Class III,through Class I, i.e. DC batteries, for a period of about 30 minutes. Thus, the contribution of batteries will be significant only in case of short term availability requirements. Apart from this Class II buses are also directly connected to the respective Class III buses.

CLASS ZV SUPPLY <6.6 KV> FAILURE

A 4 Ht.

i

521 513 5241 5241 BOS GRID GENI 522 5241 BOSSUT1 СВЭ CB34 CB33 F/G/R UT1 CB16 E/D/H

FIG. 9.1 CLASS -IV SUPPLY FAILURE

5231-7

FIG. 9.2 CLASS-Ш SUPPLY FAILURE

113

Э^5.1 Reliability AnalviHw

The details of reliability analysis are included in the Fault Tree of figure ).3 and are identical to the analysis as in f 1.

IQ FIRE WATER SYSTEM

In NAPP a system of constantly pressurised and readily available water supply system has been arranged to tackle' the type of fire where wit.er can be effective. The ¿yatem comprises storage of water, pumps and piping network terminating with hydrants and sprinklers at various locations in the plant premises. Besides this, fire water system acts as an emergency backup toa. Feed water systemt in the event of auxiliary boiler feed

water system failureb. Active process water andc. Process water cooling system

1Q..J—Syflten .Dfig.crlptlpn

The main source of fire water is the storage available in the natural draft cooling tower basins and the cooling water tunnel connecting the basins with the cooling water pump house. The fire water pumps are common for both units 1 and 2. One electric motor driven pump and three dedicated diesel engine driven pumps have been provided for this purpose.

10.2 Reliability AnalygiaReliability analysis is done for the on demand, failure

probability of fire water system as backup system to process systems with the following assumptions.

i. Two pumps out of three diesel engine driven pumps should be available.

114il. At a time backup ayates la required for one unit only

iii. Piping failure in any part of the fire water ayates, if not iaolated, can affect the aupply to the proceaa ayatesa.

Total pipe failure of thia ayates and the failure of the puspa ( two) are the contributing factora for the fire water ayates unavailability. The detaila of the analyaia are shown in the fault tree(Figure 10.1). The failure data used for the reliability calculationa is given in table 10.1. The calculated value of desand failure probability of thia ayates ia 1.0*10‘3/D

TABLE 10.1 : FAILURE RATE DATA

S .No Conponent Failure Rate

1 Pump 1.0*10'3/Da

2 Circuit Breaker 1.0*10*J/D3 Check Valve 1.0*10’4/D4 Pressure Switch 1.0*10'4/d5 Manual Valve 1.0*10"4/D6 Diesel Engine 3.0*10°/D7 Pipe Rupture 3.0*10'5/sec-yr

(dia >3") / 2.0*10‘5/ft-yr

äVlOOl CH. V.

<55*10-4-PIPING MS

VALVES

JL 3xio“4иCCF DE* s

7141 P2 7141 P3

fire baserS7STBIS7

10r3/d

PUMPS1?1X10-4-

CCFPOMPS

DOMESTICBASER

7141 P4 7141 P9

-éxio*

-¿r¿r ¿ ¿T-ó'1'"TT Ó-* ó <r<r¿r¿r <f

BBCIRCO— PS POWER POMP Viooe RECIRCO- CB DE PS POMP V1032V1002 CB ___ г л'т'тлм г.тмв тгАТТ.пиК FAILURE CH. V.

LATION LINE SUPPLYPOMP

FAILUREVlOOe RECIRCO- CB DE PS PUMP

LÄSION LINE FAILURE FAILURE

FIG. ю.1 FIRE WATER SYSTEM FAILURE

11711 REACTOR SHUTDOWN ЗТГГЕМЗ

The reactor shutdown system(RSS) is designed to automatically shutdown the reactor to prevent any damage to the plant which might subsequently lead to the release of radioactivity. It is Imperative that the RSS minimised the probability of failure of both the fuel structure and the primary system boundar/ under various conditions of operation, transients and the various postulated accident conditions. In order to achieve this, redundancies and diversities are incorporated into the design. To a large extent,, every initiating event is monitored using diverse process parameters so that the reactor shutdown function is not impaired even in the event of common cause failure of the redundant units.

.H-Л Sygtea-SsgcyiptignThe reactor shutdown system comprises of:

13.1.1 Instrumentât J^r.Process monitoring Instrumentation is used to monitor the

various process parameters like pressure, temperature, flow, radiation etc. In general, all the state variables which depict the operating environment of the fuel integrity and PHT pressure boundary, are instrumented to generate signals to be used by RSS. Limits on these parameters are so decided that under any abnormal conditions, no damage occurs taking into account the severest effect of CCF of RSS insturmentation. All the trip parameters are categorised into a) Absolute and b) Conditional trips and are arranged in triplicated channels.

11.1.2 TllF-tofllgTrip logic processes the information received from instrument channels, performs the necessary logic using 2 out of 3 coincidence scheme and provides signal to the clutch coils of the primary shutdown system( also called mechanical shutoff sysmtem or MSS). The function is performed by relay logic which operates on 48V D.C. Three seperate and independent sources of

power are used for the three channels, in view of the single failure criteria. Fourteen shutoff rods are divided into two groups of seven rods, each group of clutch coils fed from seperate 90V D.C. sources with a backup.

..1.1.. 1.«3 Shutdown devlcci

Shutdown devices ultimately trip the rector by introducing adequate amount of negative reactivity. The system has diverse, redundant provisions in the form of a. Mechanical Shutoff cods

This comprises vertical tubular cadmium rod .elements distrubuted in fourteen locations over the entire core with independent winch type drive mechanism for rod. These rods are held parked on the top of the core with the help of rope drum, electromagnetic clutch and irreversible worm and worm wheel drive. Upon receipt of shutdown signal the electromagnetic clutches are deenergised and the shutdown rods fall by gravity to bring about a quick reactor shutdown. Compression springs installed on the shutdown rods ensure the initial acceleration and a dashpot assembly absorbs the kinetic energy at the end of travel. A single failure in the primary shutdown system would not constitute a system failure under all conditions, b. Liquid poison injection

The system comprises of 12 tubes passing through the core and is divided into four banks of three rods each. Each bank has an associated poison (borated D20) tank and a Helium pressure tank and is independent of the others. The high pressure gas tank(TK-6) is connected to the liquid poison tank(TK-4) through fast acting solenoid valves. In order to reduce the probability of spurious injection, two SVs are connected in series. These valves are normally closed and opened whenever the shutdown system is required to act. All other SVs in the pressure balancing line or the Helium recirculation line are open during normal operation and closed on demand. Since three of the four banks provide sufficient reactivity depth for the reactor shutdown, nonavailability of one bank is not unsafe.

119The signal for the operation of nechahical shutoff rods is

derived fron the procees nonitoring instrunentation whereas, the liquid poison injection in addition to selective process parameters is also actuated when two or more shutoff rods fail to enter the calandria in a stipulated tine after the reactor trip. The two shutdown nechanisns aré independent and based on diverse node of operation and thus, are not anenable to CCF.

Safety Analysis

The RSS conprises of a nunber of trip paraneters which are actuated, depending upon the nature of the initiating eyent or the fault situation. However, it can be assuned that at least two parameters will be actuated for every initiating event. The details of the safety analysis are shown in the fault tree(Figure 11.1). The CCF block in the instrumentation includes common causes affecting both the paraneters because of whoch a low value of 0( p«\CMF/A«0.01} has been assuned. The safetyanalysis is carried out on the basis of the following assumptions.a. Test interval is fortnightlyb. Unsafe failure rate of the channel is 1.0*10's per hourc. Short failure of the switching diodes in the trip logic is

not unsafed. CCF of trip relays is due to welding of the contacts in the

ladder network for which a 0 of 0.01 has been assuned.e. Based on CIRUS experience[15], the probability of failure for

a shutoff rod is 2.0*10~5 per demand which for the present analysis, is assumed as 6.0*10~s per denand to account for the design differences.

The MSS reliability would then be 14C2 (6*10“5 )2 or 3.3*10’7 per denand. Hovever, taking into account all types of failures the reliability of nechanical shutdown rods have been taken as less than 10’4 per denand.

The details of the safety analysis of the liquid poison system are as shown in the fault tree(Figure 11.1). The analysis is critically dependent upon the assumption that one bank is redundant since the two nain injection valves are in series and any one of these failing to open would lead to failure of the poison injection in one bank. All the other solenoid valves which change state on denand are redundant. The probability of

RSS UNSAFE FAILURE

Л2*ю' S

. D E FИи*10Г5/Йг P ' JZjuo'Vd

F16. 11.1 REACTOR SHUTSOWIN SYSTEM FAILURE /оо

§Сн

R&S SP URI COB FU LUKE

Л0*/*'•У

POZSOS 2HJECTX0B SYSTEM

■ist \

TH'tr

1S. 0.RODS

INSTRUMENTS

А O’14/tV

'T

¿r*

CLUTCH CLUTCH D E FOPES

ihr

sacxuF HAzaSUPPLY

Tr•32/Yr

90 V DC

ОЭ/ïrCLUTCHSHORT~s

тцw

DIODE !03*10 "/Hr OPES-b

FIG. 1^2

REACTOR SHVJtdown SYSTEM SPURIOUS FAILURE

121

NO POISON INJECTION

5*10• \T

42Í2*5*io"s

SV16 svi SV2 SV17FIG. 11.3SAFETY ANALYSIS OF NAPP POISON INJECTION SYSTEM

123failure of a bank ie 2.5M0'3 is, thus, governed by the series valves and that of all the banks is 4.0M0"5, because 2 out of 4 banks aust fail on clenand.Xfifit—and—Maintenance Contribution; The test contribution froa main injection valves is negligible due to series configuration. The maintenance contribution, assuming 24 hours down time for maintehance action once in six months is

^Maintenance -4*3«V<24*2)/(Í*720) "4.0*«-*

where P0 is the probability of failure of a bank.Common Cause Failures; The analysis of CCFs for secondary shutdown system is associated with the common failures of the valves in the redundant banks of the system. Since all SVs are energised and their status displayed in the control room, the probability of any operator error during test and maintenance is considered negligible. The contribution due to design and environment is shown in the fault tree(Figure 11.3).

11.3 Reliability Analysis

The reliability analysis is associated with the spurious reactor trips due to failure in the RSS. It is assumed that the accidental dropping of a single shutoff rod or poison injection in a single bank of tubes will cause a reactor trip and also, the switching of backup supply(90V) is faster' than disengaging time period for a shutoff rod clutch. The details of the reliability anlysis are shown in the fault tree(Figure 11.2)

11.4 ResultaThe probability of unsafe failure of the reactor shutdown

system is conservatively estimated as 2*10~s/demand, the majority contribution being from CCF in instrument channels.

12 EMERGENCY CORE COOLING SYSTEM

124Eaergency Cor« Cooling Syetem(ECCS) is designed to reaove

the decay heat fron the fuel following a loss of coolant accident(LOCA) and provide means of transferring decay heat to the ultimate heat sink under all credible nodes of failure of the primary heat transport 8ystem(PHTS). Two different, systems are employed, one for handling large and medium LOCA and a second system for handling Small LOCA. In this section a safety analysis of the ECCS for large and medium LOCAs is presented. Spurious injection of ECCS is, however, not considered as it. is not possible due to the presence of check valves in the ECCS lines whose opening is governed by a positive differential pressure between ECCS and PHTS(which does not exist under normal operating conditions) when the signal is spuriously actuated.

i

J2-1 Svatca .DgacrlptlonECCS consists of (a) a heavy water accumulator (b) a light

water accumulator and (c) a recirculation system and associated piping and valves. Upon the occurence of LOCA conditions as

sensed by low inlet header pressure signal and/or differential pressure signal, signal for injection is initiated. Depending upon whether the injection is type I or II or til, the appropriate valves are operated and heavy water injection takes place. As soon as the heavy water in the heavy water accumulator gets exhausted and the system pressure falls below 32 Kg/Cm2, light water tank gets pressurised and provides core cooling after rupturing the rupture disk which normally isolates the light water tank from the PHTS. After water in this tank gets exhausted, as sensed by the low level sensor, the two out of four recirculation pumps( which are already started when the LOCA signal is generated) take suction initially from an overhead storage tank and later from the supression pool and cool the core. In type I injection depressurisation of PHTS is done during light water injection/recirculation mode by opening MV~.38 and MV-39. Prior to light water recirculation, the recirculation pump discharge passes through MV-52 back to pump suction. However, during recirculation MV-52 is closed.

12^2_Reliability Analvnla125

In carrying the reliability analysis, the following assumptions were made.

1. Only one of the two sets of (Э each) pressure sensors located on the two inlet headers is considered, as only one set is actuated depending upon the break location. Several seconds would elapse before the other set is actuated.

2. Only one of the two sets of (3 each) differential presauretransmitters between the inlet and outlet headers isconsidered as only one set is actuated depending upon the break location.

3. Pressure relief valves located in the gas lines of thepressurizing tanks for heavy water and light wateraccumulators are not considered in the analysis, as the gas pressure in the pressurizing tanks is monitored all the time.

4. In the recirculation loop P2 is considered as a backup for P1 and P4 is considered as a back up for P3 eventhough any of the pumps(excluding P1) can act as a back up for any other pump.

5. A monthly checking time is assumed. However, theinstrumentation located on the headers, which is normally not accessible, is assumed to be tested once in six months. Based on the above assumptions, a fault tree for the ECCS

failure has been constructed for type I and type II injections(figure 12.1 and 12.2). Since type III injection is similar to type II injection, the fault tree drawn for type II is also applicable for type III.

As the valves under test condition are provided with an override when LOCA occurs, testing does not contribute to ECCS inavailability. The failure data that was used in carrying out his analysis is shown in table 12.1. The data for components ike motor operated valves includes contribution of the actuator ircuit as well.

From the fault trees, shown in figures 12.1 and 12.2, the rrobability of failure of the ECCS system on demand during any /ре of injection is 3.5*10 3/demand.

126

«

J2,3 bona Term Operation of ECCS Pumpa

Since the demand on the continuous operation of ECCS is envisaged, in the event of a large or medium LOCA, for at least a period of two months, the long term reliability requirements of the ECCS recirculation system must be ensured. For the analysis of long term operation ECCS pumps are considered at a four unit system with one operating and three standby. Operation of one ECC pump is considered to be adequate during this period. It is also assumed in the analysis that there are no pumps under maintenance at the beginning of long term operation. This is justified because the most probable state of operation of the system is the operation of the first and the second of the four pumps during the initial period. Five cases (to study the effect of break down maintenance duration,if necessary) as detailed in table 12.2 are considered.

A computer program using Markov Approach [9], that was developed,for continuously operating systems, is used for carrying^out this analysis. The five casesCshown in Table 12.2) are analysed using the program and the results are shown in the same table. Since the pumps are located in the annular region, they are not subjected to extreme environment. However to take cognisance of the stresses due to operation for longer duration the 95th percentile value of the pump failure rate as given in reference [1] is used. From the results of the analysis( figure 12.Э) it can be seen that thé probability of failure for long term decay heat removal with no repair of ECC pumps for a mission time of two months would be 2*10~* per mission with a pump failure rate of 2*10~4/hr and 1*10° per mission with a pump failure rate of 3.0*10'4/hr. To illustrate the sensitivity of the unreliability to the failure rate of the pump during operation, calculations for a set of failure rates have been done and the results are shown in fig.12.3.

Л2.Л, Соадоп.Сйиас FflilureaRedundant óyeteme are susceptible to common cause failures

due to cosnonness in designf operating conditions, environnent, test and naintenance and human error. As indications/alaras are provided to monitor the status of systens/conponents inportant to safety, the likelihood of CCFs is reduced during test and naintenance. In instrumentation common errors due to improper calibration are not out of place here. The provision for physical separation of systems, like pumps is essential to ensure that malfunctioning of one pump does not affect the performance of the other pumps. In this analysis common cause contributions are conservatively estimated to see the susceptibility of the system to CCFs. As can be seen from the fault trees the CCFs considered are in 1)instrumentation 2)valves 3)pumps and 4)strainers(choking due to inadequate quality or peeling of paints on suppression pool liner)..

128

Table 12.1:Failure Rate Data

sNo.

Conponent Mode FailureRate(per hour)

Probability of Failure on Demand

1 Pressure/Level 1.0M0’6 1.0*10”4Transmitter,DifferentialTransmitter •

2 Indicating Alarm 2.0*10”®Meters or FIA All modes 4.0*10”®

Incipient 0.648*10”®3 Motorised Valve 3.5*10”?4 Solenoid Valve i.o*io”t5 Check valve Failure to Opei i - i.o*io”:

Leakage 3.0*10’7 1.0*10"46 Relay 1.0*10”47 Time Delay Relay Premature

Transfer 1.0*10”4Fails to

Transfer 6.0*10”®8 i.Failure of NC

Contact by Opening 1.0*10”7 4.0*10”5given not energis« idii.Short acrossNo/NC Contact 1.0*10”® 4.0*10"®ill.Failure of N0Contact to close 3.0*10”7 1.0*10"4Given energised

9 Flow Transmitter No 0/P for I/P 0.258*10”®(IEEE-500) Incipient 0.053*10”®

10 Flow Element No output 0.216*10"®Incipient 0.245*10:6

11 Strainer 1.0*10'3 4.0*10 112 Pump 1.0*10"313 Circuit Breaker Spurious Trip 1.0*10 s 4.0*10”3

Failure toTransfer 1.0*10 3

14 Rupture Disc • fi _ 3(Diaphragm) 6.0*10 e 2.0*10 3

15 Relief Valve Failure to _ *Close given 2.0*10 3Open

1

129

Table 12.2:Unreliability versus Tine (For Various Repair Times)

Repair Time* 7 Hours 24 Hours 36 Hours 48 Hours NoP'::pairOperating Time!

15 Days 0.94M0"9 3.18*10"9 9.32*10"9 1.89*10*7 5.12*10"®

1 Month 1.93MO'9 7.15*10"® 2.26*10"7 5.01*10~7 7.5 8 * 10" 9

2 Months З.ЭЗМО'9 1.31*10*7 4.93*10"7 1.13*10"® 1.0U*10"3

3 Months 5.92*10"9 2.30*10"7 7.59*10"7 1.75*10"® 4.39*10"3

ECOS ТУРЕ 1 INJECTION

I 3-5* lo3/d

U-pJ

?

s ъ

DESIGN TESTS.MAINT.

<5 Ъ

DESIGN EHVIRO——NHENT

FIG. tt-l FAILURE OF ECOS (TYPE 1 INJECTION)

FIG. 1X-1 ( X ) FAILURE OF ECCS (TYPE 1 INJECTION)

CO

PT—114 DPIA-52 DFIA 52-1 Í PT—117) <.D?ZA-53> < DPI—53 —11

RZLAÏS DFIA-ie-1 DPI—18< DPIA—17—1) t DPI—17)

DPT—83 C DPT—80>

R-3T75

R—4002

* APPLICABLE T0 TYPE Л &Ж INJECTIONS ONLY

FIG. .CL-1 (Y) FAILURE OF ECCS (TYPE 1 INJECTION) COro

LEVELINDICATION

¿10-6 [---- Ti'i

MV-15. 16

~io^~F л

ГBELAYS

LT- LIS- LIS— 143 63 63-

VALVES

7\

ACTUATION

à

\_/ О- О- ъ ЛX )'—'

--1U

ra

R-3764

a-3627

a-3670

a-6261

MV 15 MV 16 IЛ

В-3671-3

8-3674

MV—3.4

~JO‘S X—Аtt

IVALVES

1ACTUATION

j

А! tЛU)

IоMV3

ъ

MV4 I~Х л л л Л л

-U KJ w' •_а- н- а- а- а-3692-33638-1 3692 3636 3865

FÏG. 12.1 (Zi 2г Zá FAILURE OF ECOS (TYPE 1 INJECTION)

R- R- R- R- R- R- R- R- R-3778—2 3778 3873-3 3653 3653-2 3668 3775 3668-23638-2

FIG. 12.1 (Z 4) FAILURE OF EGGS (TYPE 1 INJECTION)гоOk

R- R- R- R- R- R-3638 3638-1 3876 3873-3 3662 3873

R- R- R- 3900 3870-63665-3

FIG.i. 12.. 1 (24 Zs ) FAILURE OF ECCS (TYPE 1 INJECTION) a»

¿Л

6 ^^5R- R- R-3629 3784-4 3668-4

FIG. 11-K26)

MV-7.8

~10'5

VALVES

■7Г

0^ACTUATION

S

MV 7 MV8à

Ó Ó ОR-3668

R-3907

R-3665-4

---------i

OR-3827-6

FIG. 1-11.1(28)

R-Эбе5

R-3665—2

R-3670-5

R-3666-3

R-3870

R- R- R- R- 3781 3616 3623 376-4

FIG. 4.1X1 (Z 9 Zio ) FAILURE OF ECCS (TYPE 1 INJECTION) ca-4*

VALVES

T

В

6R-3976

-5~I0

ZZ1ACTUATION

ОR-3784-5

FIG^_1X1{Z11) E. С. C. S.

R-3665-5

CO00

ôMVl MV2

10-5

ЛZL

6" 5 ~5 о 5 5R- R-3900 3830

R-3630

R-3778-3

R-3775-2

R-3778

FIG. 12--1(213) E. С- C- S.

ъ

R-3775

FIG. 12/1 (Zw) FAILURE OF ECCS (TYPE 1 INJECTION)

140

MV-31 <MV134)

r_- 10-3

c5MV31

(5R-3666-4

ОR-3668

1I ACTUATION

Z

X

Ъ ■ ó 5 5 Ъ~Ъ

R- 3767-4

R- 3767

R- 3764-4

R-3784

R-3629-5

R-3629

FIG. -1rl(Zl7) E. С. C. S.

<гVZ3

LIGHT BAIES XKjBCTioa

25*«0"*

!w«>o~»!-ie*5Mv-3.6 !

À

ÖV44

l0'4 : -io*5! JJV—7»e I

* I• tÀ

diî

HV-24

IA

:ао«иг» 1ючо~*'O l*"'31

< MV—134)8S

>10'UV13 V14

OSS2CB TESTAMAINT.

DESIGN SKTZBO-

FIG. 1.12.*2 FAILURE OF ECCS (TYPE 2 INJECTION]

äOJ

[Г и1 С: L

о8-Э641

FIG. .12..2 (Z 12) FAILURE OF ECOS (TYPE 2 INJECTION)

зсн

i MV—7,В

VALVES ACTUATION

3907 3Ô53 3653-33827-63665—4 3775 3775—4 3668 3668-2

FIG. .12.2 (Zœ) FAILURE OF ECOS (TYPE 2 INJECTION) 144

MIS

SIO

N

145

о

146

13 SMALL LEAK HANDLING SYSTEM

Snail leak handling system(SLH8) is designed to senove the decay heat from the fuel following a small LOCA and provide means of transferring the decay heat to the ultimate heat sink under this type of failure of the primary heat transport(PHT) system.

■U-l-Syatem Dcacriptipn

SLHS provides for sufficient D20 transfer to PHT storage tank for making up of losses from PHT system. This system' is actuated by the low storage tank level signal when the system pressure is >55Kg/cm*. This signal causes an automatic reactor trip also. Initially about 15 tonnes of 020 is transferred to the PHT storage tank from the ECCS D20 accumulator and later, when this gets exhausted, a 00 storage tank, 3211-TK-l, loacated outside the reactor building is used for supplying the necessary 02)0 by means of the pumps,ЗЗЭ5-Р7 and P8, provided for thi4 purpose.

During the , recirculation phase, either of the vault collection pumps 3491-PI or P2 takes suction from the sump collction area and pump the 'spilled 020 to the storage after conditioning(i.e purification and heat removal) it. The pressurising pumps or the PM pumps maintain the PHT ay., zem pressure and inventory at the controller set point. This of operation can be continued until leak is identified and plugged.

11.2 Reliability AnalYflla

Reliability analysis is done for on demand failure of SLHS and the details are shown in the fault tree(Figure 13.1). The failure rate data used for reliability calculation is given in table 13.1. The calculated value of the probability of failure on demand of this system is 2.0*10 *.

TABLE i:i.1 : FAILURE RATE DATA

S.No Component Failure rate

1 Level Transmitter 1.0*10’4/D2 Pressure Switch 1.040"4/D3 Control Valv-ï 1.040“3/D4 Pimp fails to

start on Demand 1.0M0"3/D5 Circuit. Breaker 1.0*10’3/D6 Husaan error 1.0*10“3/D7 Check Valve

(he »vy reverse leakage)

3.0*10”3/D*

ß Motorised ValveFails to close 1.0*10"3/DStuck closed 1.0*10’4/D

9 Strainer-Choking 4.0*10”4/D10 Heat Exchanger 8.0*10”e/tube-yr11 Level Switch 1.0»10"4/D12 Piping Rupture 3. OMO”5 /sec-yr

(dia > 3*) 2.0*10~S/ft-yr

* Upper bound (95 percentile value)

SLHS OEMANO FAILURE

i 2^ю*г/<1

И‘2ХКГ> 11-2x10гЗRECIRCULATION

SYSTEM

АMAKEUPSYSTEM

5V"

PIPING

<Дю"3 ^Ю-а^1СГ3^4х1о-*

▼99 MV ИР STRНЕ 104 105 4

PUMPS

¿Г▼100

21X10"^ • СН. ▼. HS НЕ

]зхю*+; ю-3ó <¿>▼ 107 HP

Р7

-А

ре

Ï7Г*PUMP+

HS НЕ+СЭ

ACTUATIONLOGIC

X’io'-5

Ó о О0 <6▼ 103 LT PS CP124 ACTU—CH.P. ATION SIGNAL

FÍG. . -13.1 S L H S DEMAND FAILURE

, RECIRCULATION j SYSTEM̂=É --z

4 ¡ I0-5 1 Î3»io-i !io"3C ó !|0-3 JS3HO"4 ho

( ] j STRAINER ; O O O

V97 !_________i HEAT EX- VSB MV2 vee LSHE Uxio“7Û CHANGER HS CH. V.

STBS МУ99

FIG. -13*1(BJ S*UHS. FAiiajm:

T?''MV94 HS HE

150и comí O.HvK' ISQLMIQH SYSTEMS

Reactor Containment ia necessary to restrict the release of radioactivity to the environment during normal as well as accidental conditions of reactor operation. Containment isolation during these conditions is achieved by closing the various inlet and exhaust paths for liquids as well as ventilation air. The reliability of the system has to be assesed to make sure that the design would meet the requirements imposed by all modes of operation.

11 ■ 1J3ygiem,-PeagiAp.tipn

NAPP reactor containment has been divided into two zones i) Primary Containment consisting of PHT system, Moderator system etc. and ii) Secondary containment consisting of Boiler Room and Dome Region, annular region between two walls of reactor building and main and emergency airlock housing. Under, normal operating conditions the atmosphere in the primary containment and in the secondary containment is maintained at a -ve pressure w.r.t. the external atmosphere with the help of ventilation exhaust fan units continuously running on Class TIT so as to avoid any ground level leakage from the reactor through the openings. Also, the pressure in the primary containment is maintained -ve w.r.to that in the secondary containment so as to avoid leakage from the Primary to the Secondary containment since the former one houses all nuclear systems.

The instrumentation logic used for; containment isolation system actuation comprises of the following triplicated monitoring channels.a) Reactor Building Pressure- This is monitored using two

differential pressuure switches and a PIA. This signal is effective only when the reactor coolant temperature at the outlet header is >101° C.

b) Reactor Building Exhaust Activity- This is monitored by Gross Gamma Monitors in the ventilation exhaust duct.

с) PHT Pressure- The PHT Pressure low signal coincident withPHT temperature >101°C.The signals from these three sets of sensors are wired in

the primary and secondary containment isolation logic circuit.In addition to the closure of dampers in the primary and

secondary containment intake and exhaust ducts, the follwing functions are governed by the containment isolation logic so as to ensure complete isolation of the radioactive atmosphere form normal atmosphere.

i) Isolation of D20 Vapor Recovery System ii) Isolation of Dryer room ventilation system

11.2.Reliability.Analygla

The details of reliability anaysis are shown in the fault trees of fig 14.1 & 2. Safety analysis deals with unsafefailures whereas spurious failures would result in the closure of a damper during normal reactor operation leading to RB pressurisation. Even though three signals could affect RB isolation, no credit is taken of these redundant parameters since it is realised that all may not be actuated for various accident situations.

The details of basic component failures resulting in the system failures are included in the fault trees. Due considerations are given to the particular modes of component failures. The failure rate data used in the analysis are shown in table 14.1.

14.3 Analysis of Common Cause Failures

The Containment Isolation System comprises of two subsystems i) Instrumentation for actuation and ii) Dampers for isolation. Redundancies have been provided in both. The actuation signals are provided by the triplicated instrumentation for

a) Primary containment Pressure high and PHT temperature>101°C

151

152b) High Activity in the Prinary Exhaust Ductc) PHT pressure low and PHT temperature >101°C.

In case of initiating events like LOCA, it is expected that all the three diverse parameters would be affected and thus i making the probability of any CC7 negligible. In case of accidental situation( e.g. Fuel Handling Accidents) leading to high activity, Containment Isolation would be affected by b) only. Activity monitoring is based on GM Counters wherein diagnostic systems are provided to monitor the performance of radiation monitors. This would reduce the duration of.unsafe failures significantly and hence the contribution of CCF too. In case of Ventilation Dampers, the majority of failure modes( e.g. Solenoid, Air Failures etc.) are safe, and spring failures only would be unsafe. The contribution to CCF would again be insignificant. However, an overall CCF contribution of 1*10~4/d has been assumed.

TABI.E 14.1: -FAILURE RATE DATA

COMEÛNENT FAILURE RATE1. Solenoid Valves/MV

Failure to operate 1.0*10'J/DFailure to remain open 1.0*10‘:/D

2. Pressure Switch 1.0*10” /D3. Circuit Breaker

Failure to transfer t 1.0*10"J/DSpurious Trip 1.0*10'®/hr

1.0*10’®/hr4. Buses(all modes)3. Relays

Failure to operate 1.0*10’*/DCoil Failure open or short 1.0*10’B/hrFailure of NC contacts by)

opening) 1.0*10"7/hr6 . Time Delay Relay

(Birne.allie Type)Premature Trinsler 1.0*10"*/DFails to Transfer 6.0*10‘!/hr

7. Dampers Failure ta operate 1.0*10"3/D8. Instriimen^atjon-gsneral 1.0*10"e/hrFailure to opereta9. Indicating Alarm Meters

All modes 4.0*10 */hr 2.0*10"’/hrCatastrophic

10. RTD Element 3.0*10”6/hr11. DP Transmitter 3.0*10*®/hrAll modes

Catastrophic 1.0*10”6/hr12. CM Counter 14.*10"®/hr

5.0*10"!/hrAll modesCatastrophic

13. Temperature Transmitter 1.5*10";/D14. Activity Transmitter 2.0*10"3/D

ЯЭ C0NTAINKH3JT5ИIH tPHIMARÏ»

¿1 1-5X10’® ino"^ l1 6m

rS

DAKFEaS ANO SOLENOIDS

CCP

OX

-4X1^4

IÎ02

15*7xi er* ia-uur* Ixi0"*lCONTAINMENT ACTIVITY PHT

Ó O Ó 6RI 160 R1061 67314 QIA-2

OPS

d»tt Ï

12я0~^ ! 2л I О"3ом ♦

SOLENOIDON «

SOLENOID

7> 5

8946 R-Z

Ö<^~b

63335PA-31

63335PA-32

015

Ï

01 3491 MV Г2.002 730 DM 4, 503 7314 DM 287,28804 7172 MV 8. 5805 7172 MV 9.59bô 7312 MV 26,3507 730 MV 28,3108 730 MV 29,3309 730 MV 30,320Ю 7312 DM 29,3001 7312 DM 32,33012 7312 DM 35,36DO 7312 DM 38,39014 7314 DM IODO 7314 DM 2/4

FIG. 1*1 PRIMARY CONTAINMENT ISOLATION FAILURE СЛ

PATH A

ÏО

SB COUTAIHMEMT SPORIOOS ISOLATION :

I в.бжю-^УНг.PÄIMAHT

CONTAZBMSRT

4-4чо*/н«Г! ffffg мхшвг СОЯТДХвМЕЯТ

: Sxiq-Vht.А

i ACTO ATI ON I LOGIC !

-¡6-£xJ0%r. p

• г . .ГД.FoasR1 SUPPLY

Л\

DAMPEKS,SOLENOIDS

INTAKE ! ! XUHPS«. I frSQIAttUA;

EXHAUSTPltMPB«.> bOCCNQt»*

3-5*10PATH В i PATH C

; 2-6*io"5/ht.

»VINTAKE

A A;з-з»ю3/нг. Г I

EXHAUST I

AJ

! ACTIViry I FHT FBSSSOHS

ÀA2-3XW5/Ht.о .

' Iбоне i.;3*ioV»k

Á ^

6 ^

j 7316 DH2 7316 DH4 -^•«■SOLENOID ♦SOLENOID

Ó Ó

¿i7314 DM2 7314 IBM 7314 DM6♦SOLENOID ♦SOLENOID ♦SOLENOID

1

о 1 1 /•'-s1 I ( /1 » —/ Ó Ó ó

¡

ÔЯ1160 _ 1 OIA-2 ВМ6 R—Y 6333» 633333

Û4

F631 ?Л32

Л Л FIG. 14. 2 SPURIOUS ISOLATION OF CONTAINMENT

В1060 67314РЭВ DPS

80,

POWERSUPPL?

A

,PRESSURE i LOGIC

5KII

L11

1 1ACTI-VI TT

PRESSURE PHTPRESSURE

5

1 T INSTRUMEN­TATION

1 ACTUATIONLOGIC

CROSSCOHBIBATIOB

COMBIN­ATION

¿¡^ Á Ä Á ф

”ö7>

CB CB 1635 1656

I?-*

A^X J X L

ISST. AL

î

Ъ

BUSH

Ó O F^l 6^5

BUS J BUS E +2CB* -»2CB«

ACTIVITY PHT

5<rP

?6 <b Z ъ

CB6 CB9 CB7 CB12

Jll J12 J10

¿

5 +2c:

Ó ÓBUS J *2C

BUS E -*2C

X12 L12 X10 LIO

FIS. ±4-Z(0I PRIMARY CONTAINMENT ISOLATION FAILURE

сл03

157JA REACTOR BÜILDING C00I.ER8

The reactor building is divided into two areas a) primary containment and b) secondary containment. The primary containment is further divided into two volumes -volume V1 and V2 . Volume V1 contains all high enthalpy 020 systems which includes mainly the pump room, fuelling machine vault and pressure relief chambers. V2 consists of all areas which are normally accessible during reactor operation. These two are seperated by suppression pool and associated vent system. In the event of RB getting pressurised during an accident,’ the depressurisation is affected in three stages: i) fast depressuristion to limit the peak pressure by suppression pool which removes about 25\ of the total energy, ii) RB coolers which take about 2 hours to lower the pressure further and finally iii) filtration and controlled discharge through the stack( under favourable atmospheric conditions) which slowly normalises the primary containment.

There are six coolers each in FM vault, South and North, and another five in the pump room which are required to be operational under post accident condition to bring down the pressure in the primary containment. Half of these are normally 'ON' during the reactor operation and the remaining half come into operation on increase in pressure. Accident analysis is done considering the availability of only 50\ of these coolers. Thus, only half the number will be adequate for the successful depressurisation of RB in the stipulated period and the operational reliability of RB coolers would be adequate under the operating or accident conditions. However, each half of the coolers is associated with a DG, nonavailability of any DG will incapacitate 50\ of the coolers. The probability of failure of this mode of operation, assuming the probability of failure on demand of a RB cooler as 3.6*10~Э, would be 1.0*10 Э. Another common cause factor for RB coolers is non-active high pressure process water system used for all the coolers. However, the probability of NAHPPWS failing in short duration of about 8 hours would be negligible. Thus, the probability of failure of RB coolers is considered as 1.0*10~3/d.

15816 REFEREHCES

1. WASH-1400, Appendix III & IV(NUREG-G751014), Reactor Safety Study, USNRC,1975

2. IEEE-500, 19773. HAA-SR-MEHO-12420,Volume II, "Piqua Nuclear Power Facility

Availability Evaluation Report'4. Heat Transfer and Thermcihydraulic Studies Related to the

Moderator Heat Sink in a CANDO Plant, Roqers,J.T,Currie,T.C.

5. Reliability of Piping in LWRs, Bush,S.H, Proceedings of IAEA Symposium on Reliability Problems of Reactor Pressure Components, Vienna, October,1977

6. NAPP-1/01560/86/B/8914, Reliability Analysis of NAPPProcess, Safety and Balance of Plant Systems

7. NREP Data Base8. MTL-HDBK-217C, Handbook for Reliability Prediction of

Electronic Equipment9. MAPP Safety Report, Volume II

10. Reliability of Emergency. A.C. Power Systems at NPPs, Battle,R.E., Nuclear Safety, Volume 26,1985

11. Dhruva Pipe Rupture Analysis12. Management of Leak in Moderator System, NAPP/32000/87/B/5916

dated August,!,. 198713. MAPP Moderator System Performance Report14. RAPS/09Ó00/0M/87/S/27 from Shri D.K Banerjee, 0S,RPS15. Operating Expérience of Protection and Safety System Of

CIRUS, DAE Symposium on Power Plant Safety and Reliability

KEY I/O NUMBER WINDOW N0 CRIO FUNCTION

1 W3-20 C3 CHANNEL K ECO WATER TANK LEVO. LOW

2 W3-23 ce CHANNEL L ICC WATER TANK LEVEL LOW

3 W3-28 04 CHANNEL U ECC WATER TANK LEVEL LOW

IJ.3 W3-J0 C5.C«C4

MV71/MV79 CLOSURE INITIATED ECC TANKS EMPTY

4.5.67 W3-32E5.E4E5.E4 ECC PUMPS BFF PRESS. LOW

M W3-33 E6.E8 ECC SUMP LEVEL LOW

I. 2.3,10,II, 12.13

W3-34

C3.CBC4.D5C4Æ3C3

ECC WATER 3432 TK1/TKJ TROUBLE

1.2.3 W3-39 C5.C6C4

MV72//UV80 CLOSURE (ЫШАТШECC TANKS EMPTY

15.32 W3-41 87 DOUSING TANK EMPTY

16 W3-42 D2 ECC HEAT EXCHANGERDISCHARGE TEMP. HI

\ 17.18 ' 19.20

W3-43 B3.C2C2.C2 ECC GAS 3432 TK2 TROUBLE

KEY 1 l/Q NUMBER WINDOW NO CRIO FUNCTION

28 et 0021 ce 3432 MV75 NOT CLOSE ECC IUPRD.Cl 0329 3432 CH.K ECC 40VDC FAILURECl 0330 3432 CH.L ECC 40VDC FAILURECJ 0331 3432 СН.Ы EEC 40 VOC FAILURE

15,32 Cl 0331 W3-41 07 3432 LauiBK oousme tvik empty

Cl 0512 3432 REMOTE HS OFF. NORMAL

3 Cl 0515 W3-2B C4 3432 L23M ECC WATER TANKLVL LOW

2 et este W3—23 06 3432 L23L ECC WATER TANKLVL LOW

1 Cl 0517 WJ—20 ce 3432 L23K ECC WATER TANKLVL LOW

1.2.3 Cl 0327 W3-39 C5.C6.C4 3432 UV80 CLOS IUPNO-TXS EMPTY

1.2.3 Cl 0528 W3-30 C5.C6.C4 3432 wn CLOS IMPNO-TKS EMPTY

o,e,7 Cl 0329 W3-32E3.E4E5.E4 3432 ECC PUMP 01 FF, PRESS LO

17 Cl 0533 W3-4J B3 3432 CPI ECC AIR DRIER TROUBLE

18 Ci 0534 W3-43 C2 3432 T24 ECC GAS TANK TEMP LO20 Cl 0535 W3-4J C2 3432 P22M ECC GAS TANK PRESS LO19 Cl 0538 W3-4J C2 3432 P22K ECC GAS TANK PRESS U3

20 * Cl 0337 W3-43 C2 3432 Р22Ы ECC GAS TAMC PRESS H19 et 0338 W3-43 C2 3432 P22K ECC CAS TM« PRESS ИЮ CI 0339 W3-34 05 3432 T216 ECC «ATER HTR. TEMP. H11 Cl 0340 W3-34 C4 3432 T25M ECC «ATER TXI TEMP. H12 Cl 0541 W3-34 CS 3432 T25K ECC WAHR TK3 TEMP. H11 Cl 0542 W3-34 C4 3432 T25U ECC WATER TXI TUP. Ш12 Cl 0543 C3 3432 T23K ECC WATER TX3 TEMP. LO13 Cl 034- WJ-J4 C3 J*J2 P20 ECC VATER TK PRESS. H3 Cl 0345 W3-34 C4 3432 123Ы ECC WATER TAM( LVL LO2 Cl 034« W3-34 ce 3432 U3L CCC WATER TANK LVL Ш1 Cl 0347 W3-34 CS 3432 L23K OX WATER TANK LVL LO16 Cl 034« W3-42 02 3432 Т1Э ECC HX OUTLET TEMP. H

9 Cl 0349 W3-33 E6 3432 L12 ECC SUMP LEVEL LO8 Cl 0330 W3-33 E6 3432 L11 ECC SUMP LEVEL LO9 Cl 0331 W3-33 E6 3432 LI 2 ECC SUMP LEVEL HI8 Cl 0332 W3-33 E6 3432 LI 1 ECC SUMP LEVEL H|29 Cl 0834 03 3432 V139 OPEN P1/P2 NOT RUN

23,26 Cl 0841 EUE1 3432 HXQI ECC FLOW PATH MPRÛ.27 CI 0842 £2 3432 HX01 BYPASSED ECC IMPRO

KEY I/oNUMBER

WINDOWNUMBER CRIO FUNCTION

1

30 AI 1351 05 3432 П140 ECC PI SEAL ТЕ IP

л Al 1557 03 3432 ТТ1Д0 ECC P2 SEAL ТЕ IP

30 AI 1565 03 3432 TTI31 PMI STATOR WOO TEMP.

31 Al 1624 03 3432 TTI4I PM2 STATOR WDG TEMP.

30 AJ 1626 03 3432 TT137 PM1 UP INSIDE 6RG. TEMP.

31 Al 1627 D3 3432 17147 PM2 UP INSIDE BRG. TEMP.

30 Al 1633 D3 3432 TT138 PM1 UP OUTSID BRG. TEMP.

31 Al 1634 D3 3432 TT148 PM2 UP OUTSID BRG. TEMP.

30 Al 1635 D3 3432 TT139 PMI LWR. BRO. TEMP.

J1 Al 1636 D3 3432 TT149 PM2 LWR. BRG. TEMP.

REV. NO./ REV. NO.

DATE / DATA DESCRIPTION / REVIZIA BY / DE

DRAWN BY/ OESENAT

PREPARED BY/ INT0CMI7

VERIFIED BY/ VERIFCAT

APPROVED BY/ APROBAT

DATE/DATA

S.ENE / T.CICERONE G.LEACH , J.THOMSONLi rv'

RENEL-CENTRALA NUCCEAROELECTRI :a cernavooa

OPERATIONAL FLOWSHEETEMERGENCY CORE COOLING HP ECC & SERVICE BUILDING

NOTES:’2'«.UM. TE-207K.I..M. TE—209K.L.M, <ЯЕ SPARE RTD'» WHOSE CHANNELS ARE WIRED TO THEIR

-.MSfECRVE PANELS IN THE CONTROL ROOM (51-32в,\ ex^Rnal FRECE JACKETS ARE NOT

,'2?ÏMÎF,Î5-Y WSTALLED. ïJSJ5*0 COLLECTION SYSTEM I-FS-3J810-PI

ORIO A2. A3. B2, B3.4-™E ACTUATORS of THIS Pv. are NORMALLY DETACHED

EJCEPT DURING TESTING. TO IMPROVE THE RELLABILTT'OF THE CHECK VALVES.-FOR DETAILS OF ECC INITIATION LOGIC WHICH ACTUATES

KMCES •(COUPLEXjr AS A SUBNOTE SEE OM 3432. -INSTRUMENT TESTING FACILITIES ARE OMITTED FOR

CLARITY. SEE 1-FS-34320-P3.-ALL MVs AND PVi HAVE OPEN AND CLOSE UMIT

SWITCHES FOR INDICATION PURPOSES: IE. EMI. LIGHTS. ETC. -THE VALVES ARE SHOWN IN POISED STATE.

UNLESS OTHERWISE SPECIFIED ALL EQUIPMENT ON THIS FLOWSHECT ARE PREFIXED BY 1-3432. AND INSTRUMENTS. SOLENOID. REGULATING AND CONTROL VALVES. PREFIXED BY 1-83432

WINDOW ALARMS

DEVICE EPS POWER SUPPLY DEVICE POWER SUPPL? CLASSUV3I assoo-Pim-cBiu ÜÑ39 3S32-MCC21A/F002 «UV30 92900^U70-€nOJ UV40 ЗЭК-ИСС20СД002 DWV4I Э2вОО-Рив»-СвЮЗ WV49 5532-UCC2IA/FÜ0J tikA/42 92900>PL70-C8203 UV46 SS32-MCC20C/FD03 1.UV43 азвоо-рив-ceio* MV59 5532-MCC2I A/RA01 UUV44 92e00-Pt70-CB204 MV60 5S32-MCC20C/RA01 U

■1МЯШ-....Ы.Д-ИЯ1 WV61 (11 UV80 4S2S00-PL70-C8202 | UV62 SS32-MCC20C/RA01 11

W63 SS32-MCC2IA/RAÛ3 itMV64 SS32-MCC2CC/RA03 IIMV65 S532-MCC21A/RA04 uWV66 IIW71 ssJi-ucc2iVRMI i aUV72 SS32-MCC20C/RB0I u

VÏ47(94KPlj(g))

D,0 (RD2) ZERO AIR GAP

TANK (СШЕПОИ TANK)

TK5 cue)37L

SAMPLEPOINT

E-*«~V120

BOILER ROOM

®/ТД>, TT-204K /^\TE-205K|(COMPLEX) (^J(NOTEI) ¡

TT—204L /^4TE-20SLl>4_J(N0TE1) I

OTE-20SMI (NOTE I) J

©,©

, TT-204L 1 (COMPLEX)

i TT-204M I (COMPLEX)

ГЛТЕ-214К1 (COMPLEX) IJ(NOTEI) I

РГ-ЗМ_____(COMPLEX)

"rttoF i7m'RS3¡Pa*5ídT_i i TT-208K /~NTE-207K I

J (COMPLEX) V^/(NOTEI) I

¡„©sk.ossS',1!DRAIN TO DjO

TT-208M /^>.TE-207M I COLLECTION SYS. > I 417 (COMPLEX)^(NOTEl) | ¿-ге-ЗЗвТО-PI ,

KEY cTimíl спей ВЖ! FUNCTION12ЯЗЕ W3-I CH.K MSSVi OPENING IMPENDING

u W3-2 пл CH.K (PI/P2) HEADER PRESS LOW

3 W3-3 C1 CHANNEL К REACTOR BUILDING PRESSURE EXCEEDS Э.ЗКРо

COMPLEX W3-4 CH.L MSSVs OPENING IMPENDING*.s W3-3 Г2.П CH.L (P1/P2) HEADER PRESS. LOW

в________ 1

w3-e ClCHANNEL L REACTOR BUILDING PRESSURE EXCEEDS 3.5KPo(g)

ЕП5ЕЗ W3—7 CH.U MSSVs OPENING IMPENDING7.8 1 W3-B Г6.Г4 CH.M (P1/P2) HEADER PRESS. LOW

9 W3-9 ci CHANNa M REACTOR BUILDING PRESSURE EXCEEDS 3¿KPo

10 W3-10 01 CH.K MODERATOR LEVEL HIGH11 W3-II Cl CHANNEL K F/M ROOM(R108)TEMP.H!GH

12 W3-12 01 CHANNa K F/M ROCM(R 107)TEMP.HIGF13 W3-13 01 CH.L MODERATOR LEVa HIGH14 W3-U Dl CHANNEL L F/M ROOM(R108)TEMP.H1GHIS W3-15 Dl CHANNEL L F/U RO0U(R107)TEMP.HIGH

16 W3-1B Dl CH.U MODERATOR LEVa HIGH17 W3-17 Dt CHANNa U F/U ROOU(R108)TEMP.HIG^IB W3-1B Dl CHANNa M F/M ROOM(R1D7)TEMP.HIGI

19.20 W3—19 B1.C1 CH.K BOILER ROOM TEMP. HIGH

21.22 W3—21 F1.F3 CHANNa К (P201/P202) H.T.LOOP ISOLATION INTTUTED

23,24 W3-22 B1.C1 CH.L BOILER ROOM TEMP. HIGH

2SJ6 W3-24 F2.F3 CHANNa L (P201/P202) H.T. LOOP ISOLATION INITIATED

27J8 W3-25 Bl.CI CH.M BOILER ROOM TEMP. HIGH

29.30 W3-271

FB/4 CHANNa Ы (P201/P202) H.T. LOOP ISOLATOR MITIATED

COMPLEX 1W3-28 ODD- CIRCUIT INJECTION IMPENDINGW3-29 ECC ODO CIRCUIT BLOCKED

IW3-3I MAIN STEAM SAFETY VALVE OPEN1W3-35 MSSV INSTRUMENT AIR PRESS- LOW

32.3433.35 W3-36 C2

C3 RUPTURE DISC LEVEL LOW

COMPLEX W3-37 EVEN CIRCUIT INJECTION IMPENDINGIW3-3B ECC EVEN CIRCUIT BLOCKEDIW3-40

Ri:!7 гл : ^TT-208K TE-2Û9KI T

Cl? ICOMPlfxIv ; (NOTED r-il

CONTACT ALARMS

KEY

1i GRID FUNCTION

COMPLEX 3432 CH.K ECC 40V DC FAILURE3432 CH.L ECC 40V DC FAILURE

О 0331 3432 CH.M ECC 40V DC FAILURE

COMPLEX Cl 0332 W3-1 3432 CH.K MSSV OPENING IMPEND-

COMPLEX Cl 0333 W3-4 3432 CH.L MSSV OPENING IMPEND.

COMPLEX W3-7 3432 CH.M MSSV OPENING IMPEND.'21.22 W3-21 F1.F3 3432 CH.K HT LOOP ISÛL IMPEND.125.26 Cl 0336 W3-24 F2.F3 3432 CH.L HT LOOP ISOL IMPEND-¡28.30 a 0337 W3-27 F6.F4 3432 CH.M HT LOOP ISOL. IMPEND.11.2 Í D 0338 W3-2 F1.F3 3432 PI/2K CH.K HT PRESSURE LOW

¡4.5 Cl 0339 W3-5 F2.F3 3432 P1/2L CH.L HT PRESSURE LOW

(7.6 CI 0340 W3-8 F6.F4 3432 P1/2M CH.M HT PRESSURE LOW

Ь Cl 0341 W3-3 Cl 3432 P3K RB PRESSURE>3.5KPq-6 Cl 0342 W3-6 Cl 3432 P3L RB PRESSUR£>3.5KPo

'9 Cl 0343 W3-9 Cl 3432 P3U RB PRESSURE>3.5KPo1--i D 0344 W3-3B 3432 ECC EVEN CCT BLOCKED

Cl 0345 W3—40 3432 HT INTERCONNECT VALVES FAIL

a 0346 W3-31 3432 MSSV OPEN

Cl 0347 W3-20 3432 ECC OOD CCT BLOCKEDTO ci 0348 W3-10 DI 3432 L203K MODERATOR L£VEL HIGH

13 Cl 034fl W3-13 DI 3432 U03L MODERATOR LEVEL HIGH

16 Cl 0350 W3-16 d 3432 L203M MODERATOR L£Va HIGH

COMPLEX Cl 0513 W3-37 3432 ECC EVEN CCT INJ. IMPEND.

COMPLEX a 0514 W3-2B 3432 ECC ООО CCT INJ. IMPEND.

27.28 О 051B W3-25 B1.CI 3432 CH.M BOILER ROOM TEMP. HIGH

23.24 O 0519 W3-22 BI.CI 3432 CHJ. BOILER ROOM TEMP. HICK

19.20 Cl 0520 W3-19 81,Cl 3432 CHJ< BOILER ROOM TEMP. HIGH

IB Cl 0521 W3-1B 01 3432 CH.M F/U ROOM fR107) TEMP. HIGH

13 Cl 0522 W3-15 DI 3432 CH.L F/M ROOM (R107) TEMP. HIGH

12 d 0523 W3-12 DI 3432 CH.K F/M ROOM ÍR1071 TEMP. HIGH

17 0 0524 W3-17 01 3432 CH.M F/M ROOM (R108) TEMP. HIGH

T4 D 0525 W3—14 DI 3432 CH.L F/M ROOM ÍR10B1 TEMP. HIGH

11 Cl 0526 W3-11I Cl 3432 CH.K F/M ROOM (R108) TEMP HIGH

O 0530 W3-35 3432 CH.M MSSV INST. AIR PRESS. LOW

CI 0531 W3-35 3432 CH.K MSSV INST. AIR PRESS LOW

CI 0532 W3-35 3432 CH.L MSSV INST. AIR PRESS LOW

32 O 0837 W3-36 C2 3432 L219K RDI LVU LOW

33 Cl 0838 W3-36 Ci 1 3432 L216K RD2 LVL LOW

34 Cl 0839 W3-36 C2 1 3432 L219M ROI LVL LOW.

35 Cl 0840 W3-36 C5 3432 L218U RD2 LVL LOW

ANALOG ALARMS

KEY I/ONUMBER

WINDOWNo. ORID FUNCTION

20 AI 1331 W3—19 CI 3432 Т21ЭК BOILER ROOM TEMP.33 AI 1361 W3-36 L “J 3432 L218K RD2 LVL LOW

35 AJ1362 W3-36 ~ C5 3432 L21BU RD2 LVL LOW

32 AI 1363 W3-36 C2 3432 L219K RDI LVL LOW

34 AI 1364 W3-36 C2 3432 L219U RDI LVL LOW

19 AI 1575 W3-19 Bl 3432 T204K BOILER ROOM TEMP.

23 AI 1616 W3-22 81 3432 T204L BOILER ROOM TEMP.

27 AI 1817 W3-25 01 3432 T204M BOILER ROOM TEMP.

11 AI 1620 W3-11 Ct 3432 T206K F/M ROOM (R108) TEMP.

14 AI 1821 W3-14 Dl 3432 T208L F/M ROOM (RI 08) TEMP.17 AI 1822 W3-17 Dl 3432 T206M F/U ROOM (R108) TEMP.

10 ГЖГШГТНП 01 3432 L203K MODERATOR LEVa

12 ижшлзяи Dl 3432 T208K F/M ROOM (RI07) TEMP.

15 Al 1631 W3-15 Dl 3432 T20BL F/M ROOM (R107) TEMP.

18 гетто гип 01 3432 T208M F/M ROOM (R107) TEMP.

13 Dl 3432 L203L MODERATOR LEVEL

24 remiragti CI 3432 T2I3L BOILER ROOM TEMP.

3 AI 2757 W3-3 CI 3432 РТЭК REAC. BLDG. PRESS

6 AI 2771 W3-6 Cl 3432 PT3L REAC. BLOG. PRESS

9 Al 2777 W3—9 CI 3432 PT3U REAC. BLDG. PRESS

16 AI 3156 W3-18 El 3432 L203M MODERATOR LEVEL

26 AI 3157 W3—25 CI 3432 T213M BOILER ROOM TEMP.

REV. NO./ REV. NO.

DATE / DATA DESCRIPTION / REVIZIA B? / DE

DRAWN BY/ DESENAT

» PREPARED B?/ NTOCMfT

VERIFIED BY/ VERinCAT

APPROVED B?/ APTO BAT

DATE/DATA

S.ENE T.CIClilONE G.LEACH . J.THOMSON-mW-í-'hi..

------------

RENEL-CENTRALA NUClEAROELECTRICA CERNAVODA

OPERATIONAL FLOWSHEETEMERGENCY CORE COOLING REACTOR BUILDING

1—FS—34320—P2 1 REV. A

В

D

CGC HT LOW PReSSURC LOOP

H.r. «AKR 331г-НП...

С/') НЖЬг гаи MAIN >саг-- TRANSPORT STSTCNi-rs-xiioo-Pipv_.j ру..е(А.С.) <А.0.>

TRW SPECIAL SATETY SYSTEM Test CIRCUITS COMRM TEST SOURCE i-rs-Mooe-pA-<$>-- <яО—é800»-Y._

PT..InjRGtI V/Í.VC

PI.. PT-u рт-гк PT-1L PT-ZL PT-IN PT-?M >T-«0l-K »т-г«-« »T-?oi-t •т-гое-L ►т-го1-и рт-гог-N3312-+CA0CR MDR 1 NOR 3 nor г H0R 6 NDR « H» 8 H0R 1 NOR з nor г HOR ft NOR 4 HD« ВPV..I PVIItl PVWl PVILI PVZLI PVIMI PV8MI PVZOIKI ругогте! PVZOILt pvzoai PV20IMI ругогн!n.Jt PVItt ругкг PV1L2 PV2L2 РУИС pv?«e ругоисг ругогкг PVZOlL? Pv20a2 pv?oi>e рчгог*еSV..I SV1KI ski ID ал 1M1 гмт гош гогмт ?01D ?cai zoiMi гогитSV..2 SVIKt гкг 1L2 аг IN? гиг гош гогкг SOIL? 202L? гоис гогиг

75I2-S„ -308V3 iS® -tf' 'И' IP -ip T -и® -w ■w -?p -Я4^ TO1-rS-MOOO-P« Ф»ID G4 ❖ÎRI0 G4 »10 C4 Ф»10 AI «t« ФjRID 0 <§>»10 G4 Ф»18 C4 <>:stD C4 »18 A* »ID А868000VALVE VIB VI7 VK v?f V3? V3I №0 VI* vra V27 V34 V33<Î> FROMI-FS-33I00-P1 75"grTd itЖ io s ILг: GtIB П an il Án St. ОGRID В 75“»18 0VA VIK V2K VIL va. VIM V2M VZOIK угоас V?0IL V202L V20IN V? огиPRV HtVIE (>**at PRV1L рта. Р«У1И PRVEN PRVROU рнугогк ravzoiL PRV208L PRVÎOIH ряугдги

mm R008 ROW ROW ROW R007 R007 ROW ROW МП non R007 R007

!;«l» tUPn.v|-

mœcVN.VCS

I « I ~l - И IПИЮЯЕДДД^ЧТВ-.» !Иь44|Я|135Д1ДЬ!ЛЕгТ!ПЕЗДИ ES3 Lr^ill íJ-?H.nКПТРШ

»м ВД кг+л ш ЕЯ ИД|К77*УДТ7ТД1Т7ГДдсдитапгцтдчтатдPtfr-»- Д fTSRiy ¿МЧУ-А L иктк. I то I ТО I IKI Ii?-S-MS0-vs I «si-уэ Г451-уг I

ЕСС WATER TANK LEVEL

cicvATi» у• 1¿r" ! "то 43KRAT0R HAIM CIRCUIT l-fX-32UD-Pl CRIO A?

CGC MODERATOR LEVEL LOOP

K L HLi-гоэ. LT-гож LT-гах LT-гознГ!.. ri-гож п-гозЕ ri-гожrcv. reveo» rev? ох reveo»

ругоз.i PV2UK1 PV20XI PV?f»lругв.э ругозкэ ругохэ ругожзSVZQI.I зугвж! SVZOXl sv?o»iSV209.3 svetoo SV20X3 SV203Q

VA уго»| veoxt V?0»lVS угожг vsox? VB03WVC угажз гекиэ VîeJHJ

" " VD угож4 V20X4 угоэм

« угожз VÎIOIS угоэмз73ie-s •SW V3 -307 V3 •304 V3

^ 1-rS-Klli-PI ф Otli п Q ИЮ n ^ ИЮ n

Ф l-VÏ-iP||»-PI ^ сто HI •O' «о нг •O »10 HT.

yps TO^ l-fS-32lli-Pt 4} ИЮ ¡7 ^ ИЮ 17 ^ И10 ni

OJUIDCtut

EGG SUMP LEVEL L-12

[мй*Зга,**143‘-¥г

NOTE1 ALL EOUIPTCNT LOCATED IN S-0I9

3-VALYE »CAKR73I2-S43C.VTГ " ÂTr * “î

I SUPPLY »

Ь

* L ИРГ-Х РТ-ЭИ

N.l SV»l SYXI SV3MI5У.г SVXÎ SV3«

пPURGE

VALVES■О—í¿-0

R/B PRESSURE LOOP

►< оVIHJ

A

-cx-

YIIBA-ХУ

EP10037

4 оloooe

ECC SUMP LEVEL L-ll

TO I _COK COOLINGi-rs-34 эго-pi“'Xn<0>-----<Ж>у-

i «г ►-

ECC PU№ DirrERENTtM. PRESSURE LOOP

vr Y«

cisk eiTK 212И 217NPI.. PT-2I?K PT-2171C РТ-212И PT-2I7ATK.. TK212K TKZtTIC TK2I2M ТК217ИPV..1 pveiexi PV217K1 pveibhi PV2I7MIPV„t PV212KZ PV?I7K2 pv2i2»e PV2I7«SV..1 SV212KI SV2I7K1 ! SV2I2NI sveiTMisv..e SY212t2 SV2I7K2 SV212H2 svzwteIV..3 1 SV212K3 SV2I7K3 j SVZI2M3 SV2I7K3AIR Stf>PLY*A' 1 73ie-S430Vfj 73I2-S430.V31 r3ie-S436.Vej 73I2-S436.V6P. PI P2 PI P2RG» S004 1 S004 SOI* SOI*MCV. *V21?* HCV2I7K *CV2I2M »CV217NPRV..I PRV?t2Ml PRV2I7CI PRV21PHI PRV2I7H1PRV..3 PffV?l?IO P»V2I71C3 P»V2I2W PRVгl7ЮVA veiZKi V21IK1 V2I2MI V2I7MIVB V212K2 V2I7I2 V2l2*e V217И?vc V?1K3 V217КЭ У21гю У217ЮVD v2tm V2I«4 veizm Y217H4VC уг1кэ V2ITK3 V2I2H3 V2I7H3vr vziacft V2l7Kft V2I2M6 V2I7И6VG V2I?K7 V2I7K7 V212M7 V217N7

ECC VALVE Ng BACK-UP SUPPLIES

PV- PVl PV2SV.. SV73 SV74TR-il ТК7ЭВ1 TK74BIПС-В2 TK73B2 TK74I2NV NV73 NV 74PRV PRV7* PRV 74V. V73 V74RV RV73 RV74VA V73BI V74B1VB V73B2 V74I2vc V71I3 V74B37S12-S- 430.Vft 43ft.V4

MV. NO/ MIC /RCV. MX OA1A uuuxiun / «им x/KCmami m/OrSCMkl тенми 'ОНЛИ X/vcnncAi ВУM.nAnici ■ > J.ncMCM

Я » M в

i - rt -MX* - piIMWORNCT COAT СОШМ7 nmi/MItT А РАОСГСТ uton

1 2 i I

A

В

H.P. GAS ISOLATION & VENT VALVES

ODD EVHN

PV PV8Ï PV83 PV82 PV84

7312-S ~ 451-V5 450-V6

PRV PRVB1 PRV82

SV SVBI 1 SV03 SVB21SVB4

PRESSURISING 8. DEPRESSURISING VALVES

RECIRCULATING VALVE RB, C-SIDE

7512-S 556-V3 517-V4

PRV PRV 35 PRV49 PRV52 PRV54

PV PV35 PV3b PV49 PV5G PV52 PV53 PV54 PV55

SV SV35 SV36 SV49 SV5G SV52 1 SVS3 SV54 SV55

* I 5 I 6 I 7[

ECC PUMP RECIRC. VALVE & LCC VENTING VALVES

x I AIR SUPPLY I

7512-S

ECC PUMPRECIRCULATING VALVE VENTING VALVES

PV PV23 PV24 PV73 PV74 PV78 PVB7 PV8B

7512-S 436-V3 430*V8 S3I-V2 517-VB 536-V4 537-V2 536-VB

PRV PRV223 PRV224 PRV173 PR VI74 PRV178 PRVB7 PRV 88

SV SV223 SV224 SV173 SV174 SV178 SV87 SV88

MP H20 TEST VALVES

_________________ PRVMR SUPPLY I--------pÇp-

75«-S

MAIN STEAM SAFETY VALVES

8 I______ 9______ I______ 10_____ I_______П

A

3614-PSV 7512-V 3611-VA 36H-VB 36U-VC 3611-VD 36П-ТК 63432-SV 63461-SV 3611-VE 63614-PS-l 36U-vr

5HL CK К V259 V016 V0J4 V015 V013 001 5*1 5*1 V077 505*1 V107

5*2. CK. L V260 V020 V0I8 V0I9 V017 002 5*2 5*2 V078 505*2 V108

5*3. Ch. H V26I V024 V022 V023 V021 003 5*3 5*3 V079 505*3 V109

5*4. Ch. К V259 V028 V026 V027 V025 004 5*4 5*4 V080 505*4 viio¿*L Ch. К V239 V032 VD30 V031 V029 005 6*1 6*1 V081 506*1 Vlll

6*2. Ch. L V260 V036 VG34 V035 V033 006 6*2 6*2 V082 506*2 VU2

6*3. Ch. H V26I V040 VC38 V039 V037 007 6*3 6*3 V083 506*3 VU3

6*4. Ch. К V259 V04 4 VC42 V043 V04I 008 6*4 6*4 V084 506*4. V114

7*L Ch. К V259 V048 VC46 V047 V045 009 7*1 7*1 VO05 507*1 VUS

7*г Ch. L V260 V052 VC50 V051 V049 010 7*2 7*2 V086 507*2 VU6

7*3. Ch. И V261 V056 VÛ54 V053 V053 on 7*3 7*3 V087 507*3 V117

7*4, Ch. L V260 V060 V058 V059 V057 012 7*4 7*4 V088 507*4 vue8*1 Ch. К V239 V064 VG62 V063 V061 013 8*1 8*1 r< s

! 'O 508*1 VU9

8*2, Ch. L V260 V068 V066 V067 V065 014 8*2 8*2 V090 308*2 vieo6*3. Ch. H V261 V072 VL70 V071 V069 015 B*3 8*3 V091 508*3 V12I

8*4, Ch. И V261 V076 VO 74 V075 V073 016 8*4 8*4 V092 508*4 V122

PV PVB PV9

RPV PRV76 RPV 75

SV SV76 SV75

V V76 V75

7512-S S36-V3 537-V4

DOUSING TANK ISOLATING VALVES DOUSING TANK ISOLATING VALVES

PV PV10 PVll

7512-S 436-V1 430-V5

VA V7B*1 V77*l

VB V78*2 V77*2

NV NV7B NV77

RV RV78 RV77

TK*1 TK70*i ТК77И1

T)C*2 TR78IÜ TK77I2

SV SV78 SV77

V V70 V77

RB. A-SIOE RB. C-SIDE

PV PV33 1 PV34 PV47 1 PV4B

7512-S 556-V2 517-V3

PRV PRV33 PRV34 PRV47 PRV48SV SV33 SV34 SV47 SV48

"i Л 2 : I 3 I 4 ¡ 5 i 6 " ! 7 ' Г 8 I 9

•cv. my •tv. MO. MIA ' pwmx / «WM ev / «

ег.и'^Я <V4PPL»5SE2flJ Р&лщжшжааж emssiài «Bi- Cfttfiuu тлхлллолшшйм апОыялшOPERATIONAL FLOWSHEET f

EMERCKHCY CORE COOUNO-MR SUPPUKS ~i - rs - 34эго - P« 1 i —10 11

66110 - PL.3 .

El ERGENCr CORE COOLING

О

оо

sc« SvirCK*

CKRRSSVIOPCNI*«IWtRDIXOOute»l/P2l

•®fb»oixRC<TDBР«Я HI

1 1

1

gglf

OiltPl/P2>4CABCBP«ss UJVCHLREACTORXK.PRESS Hi

СкЯRSSV*SCPOtiNCUPENDINGCHHPI/P27<*DCRP«ss. LOW

CHNRCACTOR «tS?HI

CkxNOOCRATORLCVCL - highDixг/и aoDx

A-b

CHXr/M ROWB-W7TOP m| "S’ CHLr/W 0004B-iOBTOP MtCHLГЛ* ROOMR-107TCNPXl

CHN"¡Rff"high

СМЛГ/N R0ONb-iosТЕМ HiChMr/M «JCB4 fl-107 temp h|

CK*IOJlCRRQDtTD4P KlCHXACCLPXJLArORTAMLCVCL LOV

»4KP20I/*»2C2»HCADCRP^3t LOWCHLBOILO)ROONtop hi

CHLACCUHLATORTAMLEVEL LOW04.ovn/psoeiKAocBPRCSS L0V

CHNBOILER BOPTC«PHLCMRACOPULATORTAMlevel low

сииU>-20l/P-202>HCABERPRCSS LOWCDD□RCUir!*ЛСГ|(У104РС*4в1«6 erguirI6M4ÜXTBLOCKED

9492-HV7Blogic inT«P QJJSCDSI*7CHAIM STEAMwerrvalve(PCM

ta Putps sirr.РЯГЯLOWcccSUMPlevelHLXOV

CCCWATER94B2-TKI/TK9TROUBLENJSVINJTBlKHTAIRp«ss low

RVPTIÄC DISC.ZERO air GARLEVEL LOWevtxCIRCUITINXC70II7PCRDIN6

CVtrtCIRCVMTlunjALTBLOCXCB9492-HVaoLoac inTRIP CLOUDSTATE

ттсвсатмсст valves f«T CLOSESDOuSPCt«MLEVO.LOW

ECC »CAT CKCHANKXDISCHARGETOP MlCCCGASЭ492-ТК2TROUBLE

OKA. OUR CLAST В MArtx ала DUX OUR 1 CLAST Вmu*a1 ̂ mf

11mmШ№1Ш ilmжII ES53EIE3EaESIE3SS3Bl®3Eai3 EsssESESEacart|3432-pi врр*аг II »««-pi ||1МГЯ.1ПАСТВ9«Х. II TEST 1 |943f-Pf ÍPP-вТ||70ТХ<ЛРСТОми I Э4зс-ре1 mr J XAXS PXESSWC PW2 | «АСГШ ХВОР«Я P*l | | «8CRATQR lEVCLL-R) f Л BO* i-lM Ц>*> 1-2041 ЯТЛСТЕТГАОС 1 ЯТ/ACSCTTaJlX I |яГТ/ЧСЯП1АХХ^ Str/CCUttAUC r/M «ай R-107 TOP T -20# ЯП/NEJTMAJLE

ЕЭЕЗЕЗИЕЯЕЗЭ^ВгЕЗЕЯЕЭ

мпдтр« «ЩТ> rvcinx«.о«ласс> снд.ссо &смссо DUCK) OutCCC) CHJ<CO

OU.OÜU ООКЖД)

СЭСПГЮ-ИХ «Urs rifCrilM.

OtMCC» OU.CCCJ C4J4CO смласс) otxdcci ocwxco C4JUTCC) C4J.CCC) DCKtCCI OUKCCCI OU.(CCC> C4J4CCC)CHjMjuD счхаии

QÉàÉkâlàlto 12i*9« sc-ei >еа u >#i-es»LOP I 3Jl2-"0l LOP 9 »9>»-M09 P«t»

•||||||||И11111|||||‘|"|||||||Г

49492-П-е«4(/-Г1-2а»( г л» «ж» t-iei п>р r/Ц «pi В-10? ÎQP.

lllipiil|îili|llil|îlll|ilili4

„t.iUA.I.iU.U,t 10 II lt>

В 1 2 3 4‘“I" ll|IHI|

n

0 2 « # В -11«94B-LI-2V

9432-TK3 VATER TRM LEVEL

]nil|llll°ini|llll|l

[iim|iiii|

9492-22 PtfP BPf. «BISS.

#1194 8*779

*9492-21-2» 9492*7X2 CAS УК РЯСП

hfaiiMPiyriо го «о w и-А*Э4зг-т!-гк 94Э2*1к9 W7B. ТК. ТОР.

Lull ni 11 il 111 им 1пм1 mil

W*e«9>liÉiMiAntiAiiiiAàt-M,о г « * • ie ;i>à

-wiiMip№.i»r

••41mliihihtiAiliAii

as s assg agi•э4з«-р|-л&./-*|-гоа.tiSIBSaggft

|Щ1|1111[1111|Ш||Ш1||||||” iinipuiiiîiiiiniiiîîiiiiiiir

г/п >д» в-107 ц»

j№Wñ.4 ifaA.J.ibÉiyiltili7.7 9 10 II 11.7СПЯ «S'^acssioc ! О

InnlimlmiliniLiitiniLi Juiiliii LuLiul

*9« 92-U-2 X9492-TK3 WATte г«як LtVtL

W4<p)^jàààA......I I 2 3 4 9 4 7 79

tttÉÉiiÉÎàiriiniО I 2 9 4 <*3492-21*2«3432*Txi/T«3 wra TX Pxtn

О Л 40 M #0 100.~|П11|П11|1М1|11ПЦ1И]"-Linliin

8 ïKp/tal00iljJJJJJJJiiîÉtÉàiL0 2 3 « 9 « 7 7.9

L

tÄÄÄ'BSSffiSTic34эе-мц pucNAwg то*

«>432-n*MI/ri-70l&$ХЖ8!ЯЯ

-îmiÎLUÎiM'WwW*p.«e>bipfeilüüliAiihiliiilihfaM

' г • 1 • -i“«9492*#l*tH/*Pl*24 LOP I 9J12-NB« Petit.loop 2 3>i2**«3B pacts. 4t«92*P|*20iM/.P1.2Q2H LlP I 3312-44Í4 РХСЛ L^y 2 39I2-N» w»t

|1111|11П|1Ш|111||“|||1ш'Г [llllllllljíllllllll¡í1|1Ш|Ш1|

J¡L.'«3432-tl-204N/-h-209«УЛ s: S:!?? ÎÎS.

*9492-M-204x/-T|-2)P BOlCX «Ш4 TCIP.

1 hiilmtidiiHliinlniiliiiABil9 -C 9 14 19A

ùàÀààïàààtk7.7^ » 10 II IL 7I

[ПП|1111|И|ф11ЦГП1|Ш1|П11|И1Г|

*9492-Ll-2»9432*7X1 VATC» TAMX LCVtL *9«92*PI*2l2x/*Pt-2l7«MKag-üar—

•PI-217« T. petitrr p*ga

M«(e>I t 1 < 9 t 1 >9

lllllllllllllll llllllllllllllll434 92-U-*4 DOUtlXt IX LCVCL

ТТЩ|11|ф[|||11||||111||111|

iLllllllllllllíilllllllllllllllL 0 20 40 *e M 100

■©/•■tooI...0 3 9 « 7 S 9 10

*9432-0-[I41/-Ll*t2tl «C0Vt27 R#*» LCVCL

7P»le>■.bitàoht.y.è.hit.it.lI 2 4 * S |0 12

69« 32-PI-16/-PI-l#isisxsatz

НемощиI *94 92-43-22X Ikoppi-gosx 7c>t|

/Z4O*94ae-«s-ШР7-204Кa *3*92-*S-20X1ДР7-21Х rcsi

/Г\O«3492-RS-21Lшрт-гоы. test «9492-RS-22L LttPT-гох TC3i|VS?

3«949e-RS-IX L0CPT-204L tcs «9492-RS-2X 1ДР7-21Х resq

(k 000JÚ j(C00Q Л*J432*RJ-2lx ■DDP7-20**» tC97 «3492-Rt-22NLOOPT-гоои тсзт{ rT^ü-LJ-i-y.[l r.PT-¿l>4 tÇST Ml 10-РО-Э TC37

FL ЗСВОТТОМ)

<__«__ В__ 10 12

!lu4 4 * IS 12

4943 LOOP 3 LOOP j

I2-PI-27/-PI-2B ! INJPRCSl M87>.M ! INJPRESS МПИМ

Or/^N

ас «croa toa

2.3

OPEN

ECC

INJECTION

VALVES

TRIGGER

CRASH

COOLDOWN

FIGURE Ó. SCHEMATIC OF LOGIC TO INTIATE (TRIGGER) LOOP ISOLATION VALVES, ECC INJECTION VALVES, AND CRASH COOLDOWN VALVES

LOOP

fPHT PRESS_ л

Яв PRESSA Л

MODERATOR

LEVEL

BLR RM TEMP.<*______________________ Л___________

F/M ROOM F/M DOOM

R-IOS R-101

TEMP. TEMP.-—-A, _ M.Л t 4'© © © РЭК I I P3L I ( P3M

-© -© T©ЙЭЙЁЙЙ Ы 0 Ф

J-

ANOО

XANOE

XHSОECCBLOCKINQHS

HSEHSU9K

ECCBLOCKINGMSHS119M

HÓ3.,

Hse HtHvMvrtrOtes

A,b •* To HP& nr>fee WÍtíAt/оЫ

¿■£6E)vJ) .¿■Ъ- i ос к /V £££e T4 en

ЯШМ. re - Tes Г ûi^cis/r To

>N¡r¡ATE LO О/С.

,L- ' ТP’s Г P>Lor LJb»rF/Q>. 1¿> £=cc V4¿.K£S

LOG/C БуЬЕГС/У.

PHI PRESSMODERATOR

LEVEL

f/M ROOMAIMTEMP.

PfM ROOMRIOTIEMP.

BOUERRM. IEMP

SkEICh No J

, *11 IRANSMII II RS COMMON10 ICC « 81« CRASH COOlOOWN STSICM

COMMON ALARM UNI! FORfccseiR вит with twoSCPARAIEOUIPUIBAU АЭ

3331MV22 J

ВP201K ( P202K) P201L) P202L Р20Ш P202M

M 3 R

SIMPUREO INTERCONNECT VALVE ISOLATION LOGIC SKETCH

Air Supply

Picure:/6 Typical HT Fressure Loop and Test Facilities

Figurs; /7 Typical R/B Pressure Loop and Test Facilities Р-ЗХ

Ч

R£*

Fí^ : pical Moderator Pressure Loop and Test Facilities L-203L

1

/. p/eeSitfÄi гита2 TñtHB PLATE.J BNb pl4T£ iif] ENÔ ГИТЕ S.5. channel riéTEÇ ¿ Connections

CAR-ñy>NG. Á44. g. ù, A íb/N <■> ЗАЯ,4. SuP^UiST/NC, tcLuny№. LÍE ТШ (-> -OE VÎ CEII. FvpT.ti.'. T IbHTE Ы i N Ç, áCt-T.

F/'eujea . £ £CC 5. HBAr exchange#, E- XPfrti $ÍN 6i ЮЕ Yj .

I )

OOl

оГ CPS ROOM»-170 ML

œ

VQLÎMCА0ЛЛГ

HSTГг u>Mt 'cwT«xP-«L ' “1 f -'“' ' "'t ,Wrt!,,:"g£?H«L-^W-w5l

kP—=©- ■*&W'

VOLtMXчяллгтtest/

c«oooo•sisar

рмкмгп*«тямиго*си•Е

■ ■у..,..г£мйЬмс^9 I

- ‘>«1____ Kt Kt&k •*■n^iVt «OcuTtn »i wri I

I -¿H-jbci-

=,.!=.=!=j.„__________ i у “^-s: --______f

r»e t*V 3/ iOOA 50m* •KV V •00* sont «rTU4.f\ ^

HhS-©« ■4

iU«M/ntv rwv>

-©Il ©,'&

PU463 (si-огв) PLI464 (SI-I7I)

(

3!

£3

“ ©-

fpLe9<SI-089)

oe««r 'CO« caaiNo»VT >4«t-*| I PU387

?Д£.- Mv , V . *>.i . I1 Ц

/пл* 'и/»« /bi/hj Ai** Ai** Av* А^Пг’Ог'Пг’Пг’Пг’Пгг ür

ГМ1 *Р*Ч |»м( sr*v tPMC

. . ca v*.vt ca valvc ca уч«х i4Si««v«i M«-yv4> >-M-«vn vcaccicr Vkrns^n?>«4t-nq

^ Ш41СА->Т>17«1ОСТ ldapi & :« ил tama.itss MTT. 0МРСКЯ ЭГМ0-4Г14 IÄA. caataa pw<l31WB PU4*ra j*o«f varen осатсв «»%t jaocr vam «catoi 43<V• »en*» re oonarc ««vu pti»»il «1JR to OPOUTC r»MVIC ама «*rc« «m>nis 0D» «aa «arn t«vH ^aa «art« ttm t*v* L»4» «userH ПШЛ1К IV4«?a» «wrc »aaatroi риItAP'CRti LU« 0«. КЛИМ ам»ce 47CH-L£/Lini

t-Г^ГЖОТТ-11 —

1 УММ DVCi TP» I

PLI3B8 / jSTaoJn»'Л#» Í«M-^TPL7(KSI-t7l>l —■ -pV 7—----^Ц-€>йГ~.... ....I N/ДЬчтеяс^т ■ >,.. ^. |И«^

,/ А*. 'im 4м i5“'“ т5 г5“'“ г5""*’ J5"'- t>w“1Q ПГвПГ‘ПГаПГ,ПГ'<>^ ^ < < <

Я /п^п ТТ ” ’’’ Оаь"

—азас.

СШ.«*УК

PLMk9(SI*0t4> ¿--- „.J

^<3>й£Г9xt-A.ran^KT

i__________________ Pi.eoi6j;s^-o^4>j

MOV 1« 90HI

игг;”-”*CC1»HCI

^^>•/•« Jcvw la/iw ^a/OT _3«/оч

,<>^Ь-<^ ^ ^ «L-jObk

MaUNCf WATCBлтт p^rI*M>I4«C

O Q Q Q

ice v^vt ccc VMvt tee v«.vx ca v^vt »♦*-*vw »«V-av»

Г esro-

ÍJfcb*PL2017 «!-0H>J

¿Zß’™•¡ьт-ощз;PUSM <«-01471

KTAIL ’A' - BflUSHLCSS CXC1TCR SYS.W^PVOLtkOC «CU>rw

,LJ HEïFT^;l,,^] LfXJ

) t»/S> nw

NOTESI V««» di>o»ik Kami «U. mncMi me v».vti (X5l¿f^S7.íi,'ElSÍ O' •«* •• «i. lemuMVm«a nutre me axtaa. vun rmrm ir iim t vauoM ano. un те. и um me i «*io opeurm mve .rmr1 ‘Ä— >fcv. » »oí uuut• . «*vk. ~t imi mk nnttctEH* Âtfflïï? Ä? ¡SffMiíSR —«• m-t M лт сом *ча* rv<irc«<i i*vv. i n»i •***.,* щм* VkCCliM OMtACflM 4M лт rm ICC 4ЧМ> МИШ

* JU.Î.*ÎAf‘irî“xÂ,2îî,;îSr ** ■ — —• ; У.*. ir¡** Ti ¡i. "ТГТ* ■и» t vi to. leu utmZZ.'bTSññZ ¿ЖГиЯГ "‘>• lia и MI o i .. и M mumm. I ntl mmm , • intu.ii■йэтпяа’М'ааглчйтг " —и -И ЧС'УК* l«PVt м «АС. w а»41 Uv'PUt «РУК.

LCCCND

i i .¿) Eö Ó ¿

"wrùwr- Sa?---тикжжппг-m/sónJit ««¿гм

1/4

[^) Ф

задЕЬ'йли«tafee

m

toeeub. porter o»

k&T&É, ti* Ы»аЕ±*шштСОЬШаШЭНMjiÊBEæafcss!

та» miBCTPt mm.

mtevin arauЛмЗгаГагнГТЯВмКВКТРЯГтягэлиг®йгтягвлнгтег

I

Г; t.í,

11

Síl■s .:*• ;:íç ■■

Ш

Ю1Е * INICRLQCKtNG W 0Л KV CMERGCKCt PQWC* SUPPL»

____ J L______ABNORMAL OPERATION (OP DG2>

ABNORMAL OPERATION (OP DG1)

LEGEND

ОФ

кг иг

сЩ!

«ст LOCK POSITION VITKUÎ КГТ

«ГГ LOCK POSITIO УИИ CCT lKSCe«C8

СО

c(©

two «CT OPtÎATQR SHtfT CXICNOCO *N0 LOCKED IM ОРСИ PQSttl»Ю KCT IKSCBTCS

eël

Ote «CT CPCRAtOR SHATf CXÎCNDCO AMD LOCKOUf WITH KCT BCtOVCD

OtC «CT (FCBAÍOt SMATT RCt«АСГС0 KCT KLD

tWO KCT OPCRAtOB SMATT BCtRACICD AMD LOCKED |H CLOSC POSlUOi ft< KEYS Afft INSCRICO.

NOTES1. LM.CSS OIHCRVISC SPCCiriCD. A(.L COU|P*CMl CM ÎHIS fLOVSKCT ARt PRCrilCP BT S?90.ALL CLCCTRJCAL COUIPKMî iNStRlPtCMIAl IOM AMD «OICCMVC RCLATII« WIU IC PRCnxt»

WltM 6K90.г. IHCRC ARC HOVIDCD ACCESS SLOCKING KCtS

TOR эг90-П8 (PLHWL SERIAL 10307 amO 5?90-DS7 <PVI465>. SERIAL Ml 15304 CO^AÂIKHTS.

X (»CRC ARC PROVIDCD ACCESS ILOCKIMi KCTSror згм-гг o>u4«m serial mi еззот and5790-11 (PUAU). SERIAL Ml 1530« IRANSTOtoCR COWARIOCNI.

unie 5 - KEY INIERLQCKING STSTCH DM 6kV EMERGENCY POWER SUPPLY BETWEEN DSli DSb ЪЪЭ, DSIOi Ю PREVENT PA^Aw.C. n" СГ D.lSll OlNERAIORS

LDI LOS LOI LOI 102 103

coe © c © о о

(T) CLOSC UC DISOMCCIIIS tWIICH. (?) a tnt uc шшмсстпс twro.

! PL1465 I ................ ■1 r

PLH66

.J..

(T) CPO »>C DS2 DISCtMCCTM: IVITCH

0 ctost uc »Яв-t WSCWCCTI« SWITCH

0 CLOSt UC DSI DISOMCOING SWITCH

ABNORMAL OPERATION - PL1465; PLM66 SUPPLIED BY DGI

PLM65•1 Г"

PL1466

X

0 DrtN UC DSI DISCDDCCIIND SWITCH

0 CLOSt UC DS9-I DISOMCCIDC SWIICH 0 CLOSt UC DS2 DISCWCCIIIC SWUCH

ABNORMAL OPERATION - PLM65; PL1466 SUPPLIED BY DG2

KEYS SERIAL NUMBER

THtCC КС» OPCDATOR PWWT DCIDACfCD A«D LOCKCD IWO LO-г КС» IKSCKTCD «HD tCLD UC LD-I KCT IS «CHJVCD

çjeo oülra« KEY OPTRA T OR SMVT CXtCNKD AMD LOCKED IN ОРЕМ POSITO»

TWEE KEY OPERATOR SHAS 1 СХТСМРСО AND LOCKED OUT. LO-1 MASTER KEY WSCRTCO AMO KLD LO-2 KEYS REMOVED

cto© e eiГОЛ KEY (FCRA10R S»WT RETRACTED AMO LOCKCD IN CLOSE POSITtCRL

6kV 0.4kV

PLI«W PLI4D6 PL70 PLB9

кетСОК

serial ml KEYсок

SERIAL Ml KCTCOK

SERIAL ml KEYсак

SERIAL Ml

К1 15302 Kl 13302 LDI Cl 150Г LOI Cl 150Г

кг 15101 кг 15301 lk Cl 149Г LO? CI 149Г

LI 153»

L2 15105

L4 - 15395

L5 - I53M

r ¡

кяжаш'сзжсягллвизтщж■siXsarsrjáZTrmmíimи ж к i if- atmüi,

орснЯгГЫГа!. rto/áHtífr/шкажг роли wmHfSm-m пгтккщ—i - rs - зетоо - иг I »ТУ.

11

I J о / ö

О

! F

TriLI-16

UNDERGROUND TANK ГК6гг.5 n3 чгг д

5290-DGI

PRIMARY FUEL FILTER

LEAK OFF COLLECTION

MANUAL PRIMING FUEL PUMP

PI07

Í-Я---*

SECONDARY FUEL FILTERS

PI-170О ENGINEINJECTION

PUMP

FRI03

P102 -Оь-1

©

©

FRIQ4

V32

V3I

%

F\

F\

V23

(T)PI-18

Tv40 XV

<£-tÿ«V45

PI-12( i

(V29

V24 STR2чх-

V25 V26-Of—oo-

V42

1Я FUEL DAY

TANK-TK2 9001

V38 V30 :: ? V28

( V27

5290-P2

V33[V37 [V35

5290-DG2 LEAK OFF COLLECTION

PRIMARY FUEL FILTER

MANUAL PRIMING FUEL PUMP

P207

■r\s- FR202

Í■Ö SECONDARY

FUEL FILTERS

PI-270

ОENGINEINJECTION

PUMP

FR203

P202 -Of-1

©

©

FR204

NOTE

DEVICE SUPPLY

5290-PI 5290-PL2019-CB20

5290-P2 5290-PL2018-CB20

1. UNLESS OTHERWISE SPECIFIED ALL EQUIPMENT AND VALVES ON THIS FLOWSHEET ARE PREFIXED BY 5292. AND ALL INSTRUMENTS. SOLENOID. REGULATING AND CONTROL VALVES. PREFIXED BY 65292.

LEGEND :

— THREE WAY VALVE (MANUAL CON TROLE D) REV. NO./

REV. NO.DATE / DATA DESCRIPTION / REWIA Of / oi -

DRAWN BY/ DESE NAT

PREPARED BY/ INTOC^T

VERIFIED BY/ VE RIF CAT

VPKWjjóbt/ DATE/DATA

И. MUNTCANUL .C. CAy^AN WT-rtRRAQ0r, P. rluÿtAV3.03.1Э

HEN eT - CENTHAtít NUpLEAHJfUltCTRIC/f q/íRNAYOOAOPERA TIONAL' -no WSHEET

EMERGENCY POWER SUPPLY-FUEL OIL SYSTEM1 - FS - 52920 - PI 1 REV. 8

тВашЗЯШ&тя >¿»-.‘í>warj4-i-itf«ani'i- .s: чг»»У»»ч.уц*4|ЖГД11.>ма «о: '«.-

10 11

t

PI-1003

Q 0FR13

□IC rr-P*SS ril ICRS

'V-

"K-

AUX. FILTER FRЧЛУДчое PDi-iooi

ОiOIL

filter

FRIt

tr'hsT Tvih

CLCAM OIRIt NIL OIL

Л--- ^

ps-e.i ps*i.i

9 01МГСЙМАС OIL »CA0CR

STentrPS-31(нгвт

rvii?

--- ï—tw

О OIL $IM>

HTRl03 <PS-3.I>

-Cökb

I с<н

tCVIM

l-rS-5?9e0-P?

-"K

i

pi-гооэО

O'

о

-v

nn—p>- AUX. FILTER FR\ггуAnte PDI-iOOl

ь о

OIL w-«etrtUCRS

OIL FILTER FR 21

'tr'«is! T«i4

aCAN OIPfT OIL OIL

ps-г.г ps-t.e9 9

INTERNAL OIL »CAKR

STB2l6r

«vue-CÿV

PS-L _<нтвгоэ!¿i0—is-э.г rO OIL SUM>

итигоз<ps-3.e>

^-4-

, iy,w icvzoi

-Ц2о-

гетгм-ф-

i-ri-sewo-рг

оспе г

LUK OIL TANK TK 02

Qjgk

r-o

кг ira POWER SUPPLY CLASSРП »n»-PL80l9-CI/l9 111 L C.P.S-PU voo-pieoie-c»/i9 Ill i E.P.S.HTRl« K9»*PLE0i9-ci/2i Ill L E.P.S.итого» эт»рсгою-с1/г1 Ml 1. E.P.S.

НОТЕ*i. otxtt qmcMisc spccinco all couipvcniAN» VMlXEX &Л THIS FLOWCCT ARC РЯСПХСО ft sm* am ALL INSTRUMCNTS. SXCMIO RCOMJin« AM COfTROL VALVES РЯСПХСО

ик»там / •V / Kssy

шл^п sxxzzj rszTT* ierra 'жил wat** i v#us

QPBTXTÏbHAirrtO IrSHSrr

wm Ш SYSnH SlitRCEMCr POWER SUPPLY

■V.

7394- ΠDP305C

(TS-106)

7394- ,DP306A <rs-l03)

7394- 0P303A

CTS-I03> .

7394-DP306B<IS-I01>

^ 7394- I DP303B <TS-10I>

DEVICE SUPPLY PCWJ

HTRI01 PL20I9-CB6

нткюг PL20I9-CB06

7394-DP305C LP53-CB27

7394-DP305E LP53-CB29

7394-DP305E LP53-CB25

GRID DEVICE ON OFF ALARM TRIPB3 TS-1004 38 °C 4SPC - -

B4 TS-100I - - 90°C -

B4 Ts-июг - - - 95°C

DEVICE SUPPLY P<W>

нтрго! PL20m-CB06 4501)

HTR202 PL2018-CB06 450U

7394-DP305D LP53-CB287394-DP305F LP53-CB30

7394-DP305B LP53-CB26

GRID DEVICE ON OFF ALARM TRIPD3 rs-2004 38°C 49^C -

E4 rs-2001 - - 90°C -

E4 rS-2002 - - - 95°C

NOTES :1. UNLESS OTHERWISE SPECIFIED ALL EQUIPMENT VALVCS ON THIS

FLOWSHEET ARE PREFIXED BY 5298. AND ALL INSTRUMENTS SO.ENOID REGULATING AND CONTROL VALVES. PREFIXED BY 6529ft 5a-CN01D

S' not Produce Ttr?p.DEVIŒ IS 1N 'EMERCf:NCY' rs-iooe » rs-гоое docs

LEGEND *E/D - ENGINE DRIVEN

REV. NO.DATE /DATA DESCRIPTION / REVU!» BY / OE

OPERA TiONA ¿-По WSHEXt*ЧЕКЕНСГ POtER SUPPLY-JACKET ГЛТП COOUNC ЮУГДг|

Ü REV. A 11 - rs - 52980 - pgR

Gl.024-amr

Оmо■>r~oo

n

Q3

O'J ç

Ш

c

d:О

о>—I 2>О>о

ЫСС 31G1 ГН-101 CR/MÍ САК VCMT FAM II GI ПИК СЯАЖ САК Vtltr ГАМ И Gl PH-U LUB Dl PRIMIMG PU*> II Gl РН-12 LUI 01 PRIMIMG PUMP 12 Gl РМ-13 JLW. СШСИАПМС Pi» Il Gl PM-K IH. CirajUIWG PIH> 12 Gl PM-13 GDCtATOR KARDC PIM’ Il Gl PM-II GCICRAIQR SCARING РШР 12 Gl HTR-U LUI 01 ICATCR II Gl HTR-12 LUI 01 ICAICR 12 G1 HTR-13 JIV. ICATCR 11 Gl HTR-14 J.V. HCATCR 12 Gl HTR-13 GOCRAIOR ICAICR G1 BC-II SATTCRT CHARGCRT24V)Gl FH-103 HAIN K VCNT ГАМ II Gl FH-104 MIN IG VCNT FAN 12 Gl AH-12 CUCTRICAL ROOM AH UNIT G1 ГН-Ш5 FIT VCNT FAN II Gl FN-10G FIT VCNT FAN 12 G1 ГН-107 FT ROOM VCNT FAN 11 G1 FN-Itt FI R0D4 VCNT FAN 12 Gl FN-IM GOCRATOR VCNT FAN II Gl АН-ll OGCR AIR HANKING UNIT G! HP-101 CLCCIRICAL CCMPRCSSQR Gl BC-II DICKL CWP. SAT. CHRGR Gl PN-ll FUCL TRANKCR PUM* Il Gl PN-12 ri£L IRAMSrCR PUM> 12 Gl HTR-16 FT RO» ICATCR CR-1 OVCBCAI CRAIC FN-SIO BASOCNT FAN FH-505 01 NANAGCNCNT CIRC AIR FAN FM-507 01 NANAGCICNI CIRC FAN FN-SOI DG CRC AIR FAN FN-303 U CIRC AIR FAN PN-51 0RA1HAGC P1M> SUMP II PH-33 DRAINMZ PIH> SIM> 12 PN-1 LUI 01 TRANKCR PUMP PH-IOI D1RTÍ Ol/FIEL TR. PlM>PN-1 VAICR ÜTC TRANS. PUNP FN-SOI MIN CRAIN HALL FAN PN-9 DRAIN Fl£L TRAMFCR PUM’

G 1.025-

2 X LO HEATERS 2 X LO PUtCS Г\

V

GOCRATOR г SCARING SPAKKaRInGINSTRUCN LÜ PUMP'S ICATCR

4HV GOCRATORGl

G 1.023—1CRAnIcaK

VCNT FAN 11-m¡m~VCNT FAN 12

G1.002

G1.026 -G1.013

__G 1.037

сне ramAUX LOADS

emm;crv LflAOSiG1.040

:n EVCN COHH G1-041 SERV1CCS

COMMON SERVICES SWITCH PANEL e523ie-PL1683FN-510 BASCICNI FANГН-ЗОЗ 01 MNAGCICNT CIRC AIR FANгн-307 au. uMucotsT one fanFN-SOI K CRC AR FANFN-303 DG CIRC AR FANph-31 drainage pum> sum1 »PN-33 DRAIMGC PIM> SUT 12 pN-i vatcr aie IRANI PUT FH-ЗП MU CHAIN HALL FAN PN-9 DRAW ЛЯ. TRAMIR PUT

V Z

5 5

i I

rvi u> ел

«S»:

CLASS 4 BREAKER BU

M

MCC 45G3 FN-301 CRAM CAK VENT FAN It G3 FN-ЗОг CRAM CAK VCNT FAN 12 G3 PH-31 LUI 01 PRIMING PUT II G3 PN-32 LUI 01 PRIHING PUT 12 G3 PH-33 l\l. CIRCULATING РИГ II G3 PH-34 1П. CIRCULATING PUNP 12 G3 PH-33 GEICRATUR BEARING Pur II G3 PH-34 GOCRATOR BEARING PUT 12 G3 HTR-31 LUI 01 ICATCR II G3 HTR-32 LUI 01 ICATCR 12 G3 HIR-ЗЭ JIV. ICATCR II G3 HTR-34 in. ICATCR 12 G3 HTR-35 GOCRATOR HCATCR G3 BC-31 BATTERY CHARGCR(24V)G3 FN-303 MU DG VCNT FAN II G3 FN-304 MU К VCNT FAN 12 G3 AN-32 ELECTRICAL RO» AH UNIT G3 ГН-ЭП FDT VCNT FAN II G3 ГН-Э06 FDT VCNT FAN 12 G3 FH-307 FT RO» VCNT FAN II G3 FN-308 FT RO» VCNT FAN 12 G3 ГМ-309 GOCRATOR VCNT FAN I 2 G3 AH-31 DGCR AIR HANKING UNIT G3 IC-301 aaniCAL cutressor G3 BC-31 OIEKL cor. BAT. CHRGR G3 PH-31 rua TRANKCR PUT II G3 PN-32 FUEL TRANKCR PUMP 12 G3 HTR-34 FT ram HEATER

S

5

§

4 Í G3.023—J

10 MC ВЯЕАКЕИ

(TD DOD SUS)see ova M4iot3-«>3

ira sus no. i)

LEGLHQ

81 - rOEOUDCr RELAY59 - OVERVOLTAGE RELAY27 - LNOCRVOLYAGC RELAY32 • REVERSE ROVER RELAY40 - HELD TAILURE RELAY51V - VOLTAGE RESTRAINED OVERCURREHT RELAY46 • xgattve seouence cwrent RELAY49 - GCKRATOR THERNAL RELAY87G - GENERATOR DirrERENTIAL RELAY64S - GCNERATQT RESTRICTED GROUND EAULT RELAY64 - GENERATOR STATOR GROUND EAULT RELAYV - VOLTHETER

0 TRANSDUCER

9 TEST TERMINAL

OE^ALSTHOMTOROwtq п^ТДГ,70*Т*..Л».М *•fiív¿'.,"*n'r P«ir"o ÍSS 1*«»ЯилГк>И о so« *P»>*ov«i а*» euat Г»A» COMWIBSlOMf р Q

GEC ALSTHCM INT'L CANADA INC.

AS BUILTJUN 1 5 1994

ОЯПК*-F-Fш.«жашизис=асг!,:.|Д.|а:1*и:;! *■ ■■H 11 !■■ 1И—I и il/ li<|i I < ■■■гдгдди1 ■—i .i < iiii i <гу-об-чз19-06-43 Î* I CAD gçr M013^0<

OEC A LS TH OMM5 0«C. 9ютл K9oe к ootionor /__

çfg /ЦДМ1 hVMMVLOHJDA HC б АО) РЯДОЯГТ *ДГчТ«170 ftaS'lD (Я ВДПШЛ^ USO ОК£СП.т

OR K*ttV * W tir ОСЖИОПЧ (0 ПО) *»»СК5П.

AECL - CERNAVQDA NPP1ттяг

DIESEL GENERATOR Nos. 2 AND 4 CLASS III STANDBY DIESEL GENERATOR SINGLE LINE DIAGRAM

Itvb Aí!4imi-¿04 iMIt ЯРI I TW VIS TTVS

A

5323-BUG 6Kv CLASS ill

i •

NOTES*I.IM.CSS OTXRVISC SPCCirrCB. *4.1. COUIPXNT

о* »ни rvovsxcr a*c ntrms и згзо.AU CLCCTRICAt COUIPXNf INSIRUXNrAVllX AND PBtirCCTIYC RCLATINQ WILL К РГОПХСО vjiu ьъгэо.

5323-BUH 6Kv CLASS III

нее - eu»

J9»CSCLCCICDSVIICM

wMANUKSTNC.

SCOPCrr

JT>Cc»cc«BCLAT

mam tic lecAitce ro» ocpcckv

© О 0 'ó Ó Ô Q|п-г«-1 л-етг-1 ji-ew*iKVA« VVMV pf-i л-г«-г п-гог-еCVM КУАТГIM.-1 iijoe-t м4.-г ci-гто-е к-гоо-г I

AVR • ",/'' EXCITATIUn ili.tM

er Г.J .

ü Lr

©•

-¡Ve- ' IMIM/V

-0» j -w/v I

r '

—^4¡an ^ I«4 IM 4J4* I_______IГ,---------- -•

» { ~*ППУ I

L

Х9сгMCUTftALGffOUMSPCмзэ-пзм

è‘-©‘гw-г

-©;JL 39-10,c*^”

-a5'

аг'Q 0-6 -ó-é^ßT-e-

i L.a£

АСА^.•19

"I '

CI-401-t SI-401-a a

о-a a-n-«©e-i Ji*4oe-i л-4ог-1KVA« K VA fî H*.-I CI-400-1 11-400-1

ф ф \

Bicsa >4 CO4TB0L ват

Ki-l si-401-г ci-401-г© © ©

о-©-©-©rr-i л-«ое-ел*4«-е и-4о«-гKVA« KVAtr

©“-““?-©1ЙгV, Jwo/w

а13L-"è-

М.Ч ci-w-e и-гоо-е L.ф т ?

er 1 PTзгзэ-Р1эм

аНон ; i

AVR l KVC _l LxtltMUON .'..SICH

а r ■U .

плшетпи M«.--‘i-'“?--------- 1

@ь

t4-3I—. Г-

-о-ЗFг

стIp600/*4

. J

<э5î {~°M0/W iа

Г-

CA« THINGясгв-кяlOOA/IOSCC IDA CENT.

_ IИ> ^ ,4.4 HV A.3KV Ivw*___________ IГ,------------- .

_--1g{IM/* Ig-}»« I

l_:SDG4»CUTPM.CROMOOCзгзэ-Ptsi«

lit

a“'ee:а»aаг

a^éaaa-

a“uv 40 зг

i LРТОТССПШ PAKU___^1ьвг______ I

. J

►CMA KVICCS

чхкя tcsceieiHM

гэ JTKL CHCCK BCLAT

39 ovee VOLTAGC RCLAY

г; IXDCR VOLTAGC «CLAV

эг ecvcesc eove» rclay

40 riCLD TAILIPIC eCLAY31V VOLTAGC etSTPADCD ОУСВСДОСМГ «CLAV44 NCOATIVC SCflUOCC сишснт eCLAY49 &ОСМАТШ nCOWM. «CLAV

ero GOCRATO« DirrCBCMIIAL «LAY44S OOCSAIOR BCSTStCTCO OeOLMO ГАЦ.Т44 СОСКА ТШ ЦАТОК САПМР Г AU. Г «LAY

NOTES»I.UK.cn VWKMISC SPCCtPICB. AU. CQUIP9CN1 m mit повкст AKt p«crrxcD rr эгэо.ALL ÇUMQCM. С0ШРЮ1Г IMSTQtPCNTATIOl AMD PPenCHVt SCVATITC VlU. DC PffCTIXCD Vim С329Ф.г.асспгтск couipbcnt. ивкинсмтАгкм. RCLATC» t»m эзгэ-вън and гтснпоипмсtcvtca ««c peeriко от вззгз.

ш^аизд—:..ил.^к»»

КЕЗСЯ EKCTvï (7 ТЛЯrwjvsïTя ¿ажCLASSЛ t M t А - I

> - fs -&т/тжт.ettuunMl-JsLш n

л

\

V

и И

с»г»1»с ом т*яр гам:«.fcVr» fcVJM fcVJIР1*»0в1 PIHOO« pi • toe/

ООО*V» AV3& *«ЗП PI-1007 si-1077 PI-1003

ODO

бэгэа бэгзвÎS-I0I0 PI-1047

&згзэ «згээ n-iooe* IC-I009A

JM OOU.CÎv SCC l-rS-3?300-PIui VAlte cuntí l-PS-UJSO-Pl

-✓Ч/Ч/ч

гV—V 4-10

! *:

i'ft

I JIN

f-l«àÜL:

:&S233 ьзгэз :I|-Iooot ГС-1009Р«

WINDING IP*, oc ICC TIFS

•e e ;

'“""T" ¡ 'î ьзгЭ4• : ic-1031-----------------------------------------------

iMjccioe ; . : :• ;

§&згмXS-I044

t*OQI>CIMICfllOCC

ё*5гэт p^\ взгг» p\ *згэ* twr» озгэо ^ Р^«гэ7 P's «гзо Р\ 4згз7 ^ч5гll*l»0 Ç-^ll-1037 ^^11-|03в ^^ 11-1037 ^-^11-103* Ç^II-1033 ^^11-1034 (^ 11-1033 ^*^П-|0» ^72^

4»Х (■) pi-ieiîV.x1 1_________11 1 111

7‘ÜCl onp|p|*c CW*IM PUMP ТАРРСГ DMAIMS

4_1 V«(.«PCC DMA IMSDA0OIMCiMicmooc

ЙГ"

£,"‘

IL.

TUQIO STAAT |Ki «A*

RCLAT J VKVU *

D*A|MTA«

О

I

! Г--'»'‘H f

_______________ ___ ! бэеэзT i PI-IOUA

^ l MAIM Aie

A|« Г OR SIAAFIM3su |-г*-эгзэо-Р1

ItT-j

»AARIICJMTCRUO»Й

Г'iL.

HADO SfAAIIMG

XT'-Sh

RUAT iVACTt •

? ?33«7-IS-104*A 43f37-IS-1 04г1

DUAL SKIN retí UCMAGC ACfURM TO IUL* TAMM TX9stt ws-згэго-Р!

Ï4 »AASurrut ГОА T/c MAROC* AIR ASSISTsee i-rt-зг330-p

Л A LUSC on IH.CT ■p-Qfw'Sp-]-— l-PS-3?370-P|

Í в !

/чу^учГии on 1Ч.СТset 1-гх-эгэго-Р1

■e ..j

LEGCND.

SCMI OOTAOTMAMO PRIMIMG I rs-20731 Vy

бэгэ7 .rs-2073i'

rCZ53-,1 5íJ7e-Hie||l-—_ — - —-i »ммП I PI-IOIII

MA|M A|R VA4.VC

r;5o-----------

AIR Г0Я STAAT!»«see i-rs-згззв-Р!

rtvy^-]1 згэо-мгпг 1Cis-I04í*>

I________згэ»-Г.»_________I

и-

t &

■TL

:S)~-

NOTO

— - rua on------ - JACKtT VATER

•— - SÎAAttNG AIR •— - LUBC 0П.

- ASItr AIR— - exMAusT gascs

• COMIUSrnM AIR

••• - LT. COOLING VAICR

L IMLCSS OTKRVfSC SPCCPICD AIL COUIPTCNT ANO VALVES W IMS n.OVS>CCÎ ARC PRtrUCO IT »31 AND ALL INSTRUMENTS. SOLCKHÜ RCGULAIfNG AND CCNfRa VALVCS. РЯСПХЕО SV 63R3L

LU» OIL OJTLCt SEC l-rs-SRSTD-PI

i_.^---------лу\у

•——cÆl---'

10

tfV. NOy •TV. MO.

DAIS / ott* oc icpp iw» / eoAm/Л •»/oc

QMMMI ГГ/ M90MI rvr^Mto •»/ Vtnnu m/ ¿&+0«IW1CAI y ft+mмвв,/ a*iKJDMAT^r /. y«>Au H па#^ P n

лета

DIESEL СЕНКЯЛТОЯ COMBINED PM DUO. PCI~ I »CV- ~rs - згзю - pi

11

’(•»MI

1 I 2 i j i 4 5 I 6 1 7 8

MOTORS POVCR SUPPLt CLASS

erwot 5433‘мссэг-сг III

CPH40I »4ээ*»сс4б-сг III

вегг X433-MCC32«JJ III

К4г 3433-MCC4D-N| III

LEGEND.

©

ОriXHCMAQGCR

STARIIMj KJÎQR

NQTESi1. IM.CSS OTICffVISC SPCCITtCO ALL COVIPMCMTS

AMD VALVCS CM (Mis riOVSMCCI ABC PWIXCD rr Э73& ANO ALL tMSimMCMr& SOLCMJU REGULATING AND CONTROL VALVCS. ВВСПХСОIt um

£ ALL COUIRMCNT ARC РКПХСР It CORCIPGND1NG MJRKR 0Г DtCSCLS L&3 OR 4.

X ALL VALVCS ARC РЯСПХСО It CORCSWOIIC MMICR or Otesas l&J OR 4.

4. ALL IKSTRUCNTS ARC PRCHXCO »V CWCSPCMOING ICMPCR 0Г DICSCLS 1&Э OR 4.

X svzuaa. sveom AND corcsptmoing sglcmiid rm oox sv4ox* SV40KI arc ccmirolco it SIARTOC LOGIC.

6. SVZCDT^ sveosn RCSPCCTÍM3 SV4037AJV403TB ARC COITROLCO IT LOAD StQUCMCCR AND OPCN г SCC. PRIOR A lie LOAD START.

■ГГ. КОУ mpt. mo.OAfX /OAU «sanrriM / ww* vr,

e5Soш,t Mi*wie wt/■лоси* I vTNnrp m/ I1 1 %

^ 1. РОСАМ/ v-T«/ vt□ гтттазт

10

ASLOriRXrtONAtbJZairSHSFÏ i,

ST AND STARTING A/R S)rs - эгз50 - яг

11

1 X

7177-V006►Ч--

1-FS-71770-P2 7177,GRID E6 V009

7131-V23I

7131-V227-:-c—О

l-FS-71310-P3 GRID CIO

. ихтз

1-PS-7 1310-P3 GRID cm

Tl-10fc9 11-1000

7131-VBSB О

u \Xj m" 1

г- 0 ©I-FS-5230O-P1

GRID C7

WATER TANKгип7

\

./ TI-3091 T T-3090

О ©

I-FS-7J3I0-РЭ GRID CIO

■O' ©TI-3009 TT-3000

HX303

l-rS-5230O-P3 GRID C7

NOTE:1. UNLCSS OTHERWISC SPCCIFIED ALL EQUIPMENTS

AND VALVES ON THIS FLOWSHEET ARE PREFIXED BY 5237. AND ALL INSTRUMENTS. SOLENOID. REGULATING AND CONTROL VALVES. PREFIXED BY 65237.

REV. NO./ REV. NO.

DATE /OAU DESCRIPTION / REVUIA 8Y / OE

DRAWN 8Г/ DCÄNAT

PREPARED ВТ/ MTOCUST

VERIFICO ВТ/ VER1FCAT /

“’^ЖтвТ/ DATE/DATA

bf J . ROSU 1. DOCARU р/ ГЕРЕНА94.12.27

f y/f Í - CENTMUrttaClEAR$LECTRtc\ ÇERNAYOOA

OPERA f¡ONAborto ГГ SHEET

RAW WATER COOLING SYSTEMI - FS - 58370 - PI I REV. A

8

DIESEL 6 U!L DING

.J5Z2>40д 6-ш- 40/;____ш

-Ä:

-Vnew

! ^ V ¿90 • ¿40/-б-À/ ^ g

№¿$10 ММ á-----1-П. j_ - T SUM

I//Û/S

S.237 TK<ômm

L.O

—CXh-//<907

\ \ ///P/9

VF—A yaréâ

ti1”

S2h7û ïta?--- -A

‘SÛG 4 coo/'oÿ múc/¿//íz

Гt-¿гу-ШГ.

£'W' 2QZА

-(xr^h-(Xf4J-

♦ 52Ï7OV4067/C90 г -*•

5237^üi—A-¿гзУом/^-х- —и

1 —1¿ ¿ 4= л

7/Ç9O-GZ0/- С*-Л/ ^ ^

5Zb70 V/0/4

\i5¿37 ГК 2.ft/run.

LO4X1-//ûô€

'v/O/Ô

т_А ЫОГ£А Y.S¿110

S-U/-PO/____ •

SU70/2OS ►Хл

52370/206

SùC Z coo//nj n?oc/c//e

r-«h-■'.i-------

XfXj--------

7/690

SZilOVsO/-^.^.-T“V1522* l/T/1

1 1 1I

SZW

7/c?û-6w-e-4/^^/ /¿P/3

70 /204

V52S>7 ~K2>

Q/n/rtj

L.0—1X1

v/oos//017

52370^ G- //- /0/

r Û A/oros52370^/305---- tSj—

t

SùGb coo//nf moc/cz/e

C-W-/OZr-A

-tXTxr-cxrvh

52370/306

7/4ŸO

7/£?0- 6/0/-£'/l/ ^ ^

52270 V/03_^5 г 370 У/04

^ J52370

//0/Z

ïySE____'C__^237 ГК/ '

V_________ ô/rl/nj

LO—tXh

У/Oté

h/orsô5¿ПО //05A

f

YS fi fi / coo/my moc/a/e

52370 //06

I

в

ог-

5238-ТК101(over По»)

LS-U04

II

______ I

нхюгА нхюгв

! оГТ--1

is- рт- юш 1047

16RK270 ENGINE

CNGINCDRIVCNPUHP

©

óPI-1004©

ENGINEDRIVENPUMP

|-гх-згз8о-Р] GRID ЕЗ

l-rS-52380-P4 GRID 03

<©

г

DUrr AND STANDBY

PUMP AND HCATCRS

vi30 J-|

-*\s-

CODLING MODULE

TT-1078 Г1-1079e оh VII6

VI17-o- 4rcvioi

V188^H Vlg^

VISO VI2I /TN

4>^i--- кэ-—i—0\ZSZ3—\sr~TV|22^ PMI3

V125 V|26 , HTB1‘I /<Г\cÿ]--- кэ-—r-r^^í—( t-4-

V127 РМИ

© ÔT T-1076 TI-1077

гг-ювг Ti-1083

© OTT-1095 TI-1096

© OИХ101

TT-1084 11-1085

© O

HX103

y VI08

ГТ-1080 Tl-1081© oV13I 4

в

l-rs-S2370-PI GRID C6

D

1-г5-згзво-Р4GRID Dl

V151

V7/ 1-Г$-52Э80-Р1 GRID El

MOTORS POWER SUPPLY CLASS

PM13 5433-MCC3I-A6 IIIPM14 S433-MCC3I-B2 IIIHTRI3 S433-HCC31-A3 IIIHTR14 5433-MCC31-M4 111

LEGEND:HIGH TEMP. CIRCUIT

LGV TEMP. CIRCUIT

NOTES:1. UNLESS OTHERWISE specified ALL

EQUIPMENT AND VALVES on THISflowsheet are prefixed by згэе,AND ALL INSTRUMENTS. SOLENOID. REGULATING AND CONTROL VALVES. PREFIXED BY 65238.

2. VALVE VI54 IS NORMALLY closed VITH a 3nn HOLE DRILLED THROUGH THE GATE. OPERATIONAL TLÓYTSHEEt У

GLYCOL WATER COOLING SYSTEM11 - FS - 52380 - PI REV. A

8

I

в

D

(over flow) V853-ÍXb

5238-TK201

V254note e

HEADER TANK

X «“X

ilrzizl

11 r O f-li г , VCCJ VCCO: i---------^----<]--- r-

j ....../^K /éK 1 [S L.-J .«i --•••■t

V227^HX202A( нхгогв!

16RK270 ENGINE

ENGINE , ^ DRIVEN I ' 'PUMP

IS- PT- 2010 204 7

ÔPI-2004

eCNGINCnqi\/rNpuhp"

-41 GRID E5

ФФ

e-2101.

DUTY AND STANDBY PUMP AND HEATERS

V230 H

-4Л>-

CDQLING MODULE

TT-2078 11-2079

© о

V2I7—О— ч

V2I6 V22B^-| V229^-|

V220 V22I HTBg3-cÿ]--- <ь—r-{5ZSZ3—(GV222 PM23

PM24

TCV201

© ÔГТ-2076 11-2077

TT-2002 T1-2083 wю© O

T T-2095 TI-209&

© OHX20I

TT-2004 TI-2005

© O

HX203

VV208

V2O0ЧХЪ

T T-2080 TI-2081

© O

V231 H

Я

I-FS-52370-P2 GRID C5

l-rS-52380-P4 y GRID

l-rS-523B0-P4 GRID DI

MOTORS POVER SUPPLY CLASS

PM23 5433-МСС32-Л6 IIIРИ24 5433-MCC32-B2 IIIНТЙ23 5433-MCC32-A6 IIIHTR24 5433-MCC32-L1 III

^V25l

\y7GRID El

LEGEND:HIGH TEMP CIRCUIT

LÜV TEMP. CIRCUIT

NOTES:1. UNLESS OTHERWISE SPECIFIED ALL

EQUIPMENT AND VALVES ON THIS FLOWSHEET ARE PREFIXED BY 5238. AND ALL INSTRUMENTS. SOLENOID, REGULATING AND CONTROL VALVES. PREFIXED BY 65230.

2. VALVE V254 IS NORMALLY CLOSED WITH A 3nn HOLE DRILLED TROUGH THE GATE

REV. NOy REV. NO.

DATE / OATA DESCRIPTION / REVIZIA BY / DE

REN it.

DRAWN BY/DESE NAT

PREPARED BY/INTOCUIT

VERmCO BY/ VEWCAT ■"я®/" ОЛТЕ/

OATAI. DOGARU P./n.aAíia I

i ч ■Г“ Ä—* ' Г1Ж 1 94.12.«

/CERNAVODAOPERATIONAL'FLOWSHEET /

GLYCOL WATER COOLING SYSTEMI - FS - 52380 - P2 I REV. A

8

5238-TK301(ovrr flow)

/"V,,..

D

V354V

note гд

LS *3106

HX302A нхзогв

16RK270 ENGINE

ENGINE , DRIVEN ( r ' PUMP

IS- PT- 3010 30-17

f] 0 0?

ÔPI-3004

гENGINEDRIVENPUMP

GRID ES

<«> 1-г$-5гзво-Р4GRID D3

l-rS-523BD-P4 GRID D1

"TV330

MOTORS PQVGR SUPPLY CLASS

PM33 5433-MCC45-A6 IIIPM34 5433-MCC45-B2 IIIHTR33 5433-MCC45-A3 IIIHTR34 5433-MCC45-J1 III

COOLING MÜDULE

TI-307B П-3079e оJ;v3,6

V3I7—t>b-

узгв^н узгф

V320 V321 НГЙЗЗ

-c^a--- <}-—i—ESZ5Z3—(fV322 РМЭЗ

V325 V326 нгя34-Dÿ]----<Ь--1—HN/N/4!--( ç

V327 РН34

TCV30I

© ОТ Т-3076 TI-3077

тт-зовг Ti-3083е ог!á¡n J

т т - 3095

0

V3C8ЧХУ

HX30I

TI-3096

O

ГГ-30В4 T1-3085

0 O

rI--l0B0 T1-3081

0 O

УЗЗГЧ

l-FS-52350-PI GRID E6

w

V3SI

GRID El

LEGEND:-------------- - HIGH TEMP. CIRCUIT

-------------- - LOW TEMP. CIRCUIT

NOTES:1. UNLESS OTHERWISE SPECIFIED ALL

EQUIPMENT AND VALVES ON THIS flowsheet are PREFIXED BY 523& AND ALL INSTRUMENTS. SOLENOID. Rrr,ULAT|NG AND CONTROL VALVES. PREFIXED BY 65238.

2. VAI VE V334 IS NORMALLY CLOSED VI III A 3nn HOLE DRILLED TROUGH THE GATE

REV. NO./REV. NO.

DATE /DATA DESCRIPTION / REVIZIA BY / OE

DRAWN BY/OESOUT

CU

PREPARED BY/INtOCUIT

ndi”SLVCRtntO ВГ/

VCRIFICAT

ÜJtçlSMgfc

R E N^E L^CKNTRALA-ffU£LEAROEU:CTRIfA CERNA YODA

BY/ DATE/ J DATA

94.12.27

OPERA TIONA E~FLO ITS НЕЕ ÏGLYCOL WATER COOLING SYSTEM

I REV. A1 - FS - 55380 - P3

8

VA­

NOTE *I. 1M.CSS QTHCRVISC arenco 4L соинчснт AMO VAL ves fM TfOS n.QVS»CCr A« PffCriXCO IT K3tL mi 4L WSTRUCNTS.sacMiia irccu.«r»« mi соигво. val ves.АЯСПХСО IT 65гэ&г. vALve v«3A is tcmM.LT anco

vim a jm* Ю.С muí i touch ГК CATC

э. val ve vas is rat ami uex o- «.reo.то PRCPARC SQ.UTOL

TOTQVS POVte SCPPLT CLASSp*n 54ЭЭ-ИССЭ1 >L4 IIIPH? IIIртэ Э4ЭЗ'ЙСС46-А6 IIIPH44 S433*>CC46.|? IIIHf*4J ■.433-*CC46-AJ 111НГР44 Л4ЭЗ-»СС46.Л III

•or. N0/nv. wo.

04 IT /041» •остлом / t» /

10

л ш è ш ь • \OPEMVKBiL riOWSHSKTGLYCOL WATER COOUNG SYSTl

1 - rs - згзво - 1 »CV11

pt-»ой

VII5

X-

cГСУЮЗ

1-Г9-52Э80-Р4

/ \/ N / \/ \

HX-401

\ / \ /\ /\ /

■O'-1075

■e TT-1074

TT-1072

1^7> [fâ

rci

HAND OPERAICD □IL PRIMING PUMP

=n^-PUVII?

-X-VIIO -X-VIII

E/D OIL PUMPS

ES-193A0 рн\г_чДз ‘ <i—TR)—[/1——Г~

VI07 VI08 ¿____ VI09

rs-l93B

LUBE OIL PRIMINGpijhps

Zho-hV104 V103

-D^O-VI06

РИИduplexriLTCRS

ГТ-1070 Г1-1071

VI32

s4»s

f1 iPT-\a96

HIW{t -j iLO

1 PI-1097

9 Ф .VI33

-r\s-TLCXCONN.ГГ2

COOLING MODULE

GENERATOR BEARING LUBE GIL SYSTEMi--------------------------------------

I LS-1113

-1112

IIII IS-52390-P3

'.RID B2

DRAIN

16RK270ENGINE

l-ES-52390-P5 GRID B&

-<5>-X — v viol

OIL FILLNORMALLY DISCONNECTED

!-Jvil3

II

HTRII

TS-I042B

HTR12

—©

—e

—о —©

PS-1003

PS-1004

PI-1001

PT-1050

TS-1042A

MOTORS POWER SUPPLY CLASS

PMH 5433-KCC3I-A4 III

PMI2 5433-MCC3I-A3 MlHTRII 5433-MCC3I-AI III

HTR12 5433-MCC3I-A2 III

5231-PM15 5433-MCC3I-B3 III523I-PM16 5433-MCC31-B4 III

NOTEI. UNLESS OTHERWISE specified ALL

EQUIPMENT AND VALVES ON THIS FLOWSHEET ARE PREFIXED BY 5239, AND ALL INSTRUMENTS. SOLENOIDl REGULATING AND CONTROL VALVES. PREFIXED BY 65839.

LEGENDFLOW GLASS

REV. NO./ REV. NO.

DATE / DATA DESCRIPTION / RÇWA 8Y

DRAWN by/ OESENAT

PREPARED BY/ INTOCUIT

VERIFIED «Г/ VERJFCAT ^ |АРяЬED^BY/

yf. M^TCANU 1. DOOARUA«,

R E N E L - CENTRAUXMjCLEAHbiha,STRICA cé&N*4V

OPERA TfOUAb-ltOWSHEET

LUBE OIL SYSTEMI - FS - 52390 - PI —1 REV

1

1 2 4 5 6

MAIN DIESEL VENTILATION AND COMBUSTION AIR SERVICE

1 2 4 5

7 8 9 10 11

MOTORS POVCR SU»*PLV CLASS

ПО? 54ЭЭ-НССЭ1-С6 ш

ГЭ09 S43D-KC4S-J4 ill

П0Э 54зэ-ны:з1-сз шГ10А з«зз-к:сз1-С4 ш

ГЭ03 54ЭЭ-НСС«3-СЭ IIIГЭОА Э4ЭЭ-МСС4Э-С4 шггоэ з«зэ*мссзг>сз шггоА Э4зз-мссэг-с4 IIIГАОЭ Э433-ИСС46-СЭ HIT ADA 3*ЭЗ-ИСС*0*С4 шгго9 34зэ-мссзгчэ шГАОВ 3433-HCC46-J4 IIIrioi V»33-»CC3WHI ш

ггш з4зэ-нссэг-н| III

Г 301 3433-ИСС43-С1 шГ 401 34ЭЗ-МСС40-С1 III

пог Э43з-*<сзьиг III

ггог 3433-*CC32'H2 ш

гзог 34ЭЭ*МСС43-01 ш

ГАог Э4ЭЗ-ИСС46-С1 ш

Г'

rPflo I}"2'

LEGENDr - TAN

ВГ - Г IRC OAHPCR AT - AIR fILTC« l - PNCUHATIC lOUVCR DA»*»CR

NOTC1. UNLESS OTMCRV1SC SPCCITICD ALL

tQUJWCHl ANO VALVES OH THIS rLOVSKCT ARC RRCriXCO Bt T352. AMO ALL INSTRUMENTS. SOICKIIO. RC0U.ATIN0 ANO CONTROL VALVfJ. RRCffKCP Ft 67Э5г.

i

L..

CRANK CASE EXHAUST SYSTEM

OCttwnON / М>1фл л

Шггммшзящщшнётшшкя^зтшя^яс^г^твгашшУ№Ж2Шг~ш1)&йелшлкат

К ff Г L - CSHTfUUr *4сиЛЙф1£СГ*Сл &Ot^färrVIfAL_J*6WSH8tír •

тгг\шт km АЯСЪонъиягю* smj7

в

V005 Ц.\

PI-1

о—*Ч—' V006

l-rs-751E0'P4 GRID ВЗ

-CÍ3-VDD4

О

Vi

7512-VHO

ТО 5237-PV103 Г®-к

l-FS-52370-Pl SI

cïo-VI09

65237-SV103 Ж

V101 _-<J----------СЛЗ—*■

V105

TO 5237-PVID4

l-rS-52370-PI S’

✓ 104 Г

м vilo

65237-SV104 Ж

VI02 _-О-----1Ж}—•-

vio6

ТО 5237-PV103[Щ.l-rS-52370-Pl S’

—cSa-Vlll65237-SVI05 ^пГ

VI03 т-К}---0<Н

V107

l-rS-52370-Pl SI

ТО 5237-PV106

ELчЛИ

V112

65237-SV106 VI08

V4I4V413 -т-—c>f—сЗ<ь А side

ТО LOUVERS ОГ S DG N0.4

V214V213 -г—i>—СЖЬ Д side

—вТО LOUVCRS or SDG N0.2

V3I4V313 т—t>—ок}- А side

В ТО LOUVERS ОГ SDG N0.3

v„3 V"4

—1>+—[><]- A side

VI15 V116

—1>—tXb В side TO LOUVERS 0Г SDG N0.1

W

TO 5237-PV303(El-rs-52370-Pl JÖL—Зь-Л

V309

65237-SV303

TK305

1-FS-52370-PÏ

TO 5237-PV304

[Щ-cía-

2;

V30IЬ<ь- -Ä-H

V305

V310

65237-SV304

TK306 V302

-XJ— -cïo—»-V306

l-rs-52370-Pl S’

TO 5237-PV305

EL V3ii

65237-SV305

TK307

l-rs-52370-PI

TO 5237-PV3067306 I-

*4

V

-PI-cikj-icV312 308

V303 -p

J-Xb—t><bV307

65237-SV306

V304 _

-*à-—t><bV30B

I-FS-52370-P2TO 5237-PV203 I— ----------tÄj-f

m 1 ve°9¥65237-SV203

TK205 V20I

J-о— ч>кьV205

65237-SV204

l-rs-52370-P2 S’ TO 5237-PV204 I-------------- i>k}—El 1 vs,° TK

206

V206

TO 5237-PV205Ш]I-rs-52370-P2 S’

Ä-fVlll65237-SV205

TK207 V203

-хъ- -okyV207

I-rs-52370-P2 TC

TO 5237-PV206 p-DÍJ-

V2I2

65237-SV206

TK208 V204-o—

-Ä-

V208

l-FS-52370-P2 H TO 5237-PV403 I-----------------E>k—Г

bl| ^65237-SV403

TK405

l-rS-58370-P2

TO 5237-PV404V404 r~

*4-ÄH

V4I0

65237-SV404

TK406

l-FS-52370-P2 H

TD 5237-PV4D5 |-----------------i>k}—El 1 ^

65237-SV405

TK407

Ж

l-rS-52370-P2 S’

TO 5237-PV406 Г-

<4-{Ль

V412

65237-SV406

TK408

Ж

V40I -p-Kj-—[>кь—<V405

V402 —-XJ-—t><J—<-

V406

V403 -p-хь—okh-4-V407

V404 _-X}-—Ыо—•-V408

NOTEI. UNLESS OTHERWISE SPECIFIED ALL

EQUIPMENT AND VALVES ON THIS FLOWSHEET ARE PREFIXED BY 5235. AND ALL INSTRUMENTS. SOLENOID. REGULATING AND CONTROL VALVES, PREFIXED BY 65235.

REV. NOy REV. ЫО.DATE /DATA DESCRIPTION / RCVMIA BY / DE

DRAWN BY/ I PREPARED BY/DE SENAT . 1 INTOCMIT VERIFIED BY/

VERtrjCAT v jpp^^n/ date/DATA

M.jMUNTEANU L 1. DOGARU H.' p.^li/hia

94.13.27Lc. /ik

OPERA TlONAlr-FbÓYtSHEETSDG INSTRUMENT AIR SYSTEM

-1,11 - FS - 52350 - P3 REV. A

0

R

¿LAPSED ÎÜŒT

LO

15

25

35

tí

45

50

55

55

SC

SSQ. STEPS

и

1 **

2 "J «4 **

5 **

tttt

10

11

Ш1Ш1РТ1Ж *r~t

Sheet 3 oí 1

CLASS III tefl" LOAD SÈMES, V KITE Щ DCs 0» TBS EUS

Cl.III KC's aa П 25,* '■Lighting, Cliss l, Class II \T190-Chiller û'rc. tap (3)

a.III ItCCs on TZ 27, CD 225 ?

LOSS OF i 1Г (ШЕР)___1220IV 00 M 3fÍr Ih)

4321-itu. Ond. txtr. Pimp

3432-ECC tap^3432-ECC Рщf 3411-End Shield Cooling Pimp

3411-End Shield Coding Panp

352S-F/V Pressurising tap7131-Sill Ser. Viter Pmp-1

7131-Siii See. Viter Рапр-З

7131-Sail Ser. Viter Pinop-3

7121-Sin Sot. Viter Pmp-1■лi:

LOSS OF CL IV S LOCA CMEXTS

1220 IV (1000 iV afterWSun 140 IV (9S.il IS)

Includes Solentor Ponj Motor. One Class l bit tecу issuatd dischtrged.3 PlOpS oui oí 3 required

1000 IV (Ц0 kV öfter lb) 1000 tv (210 tV after W

I1

ПЯ1.К 1

ИЕШЕЯШ CUSS III Ш LOAD SEOUMC! HIM ПЮ DCs QH THE BUS

ELAPSED TIME SEQ. STEPS Ltà DESCRIPTIF COKTBDL LOSS Of a IV (RATED) LOSS OF Cl !Y S LOCA COHMEXTS

10 12 3341-Shutdom Cooling Pop OFF-OR 4 Reset (Bot Runpl №(203! Inhibit Run an shutdom cool ¿.80 13 7134-Recirc. Cooling Ровр-1 OFF-STST-OR

r

Run 125B № (ISIS №) Run I75C (128c ■ Onlf one pump per bur to run

85 147134-Secirc. Cooling Ьвр-З 1 Reset (tot Run) f Reset (Rot Run) 4

85 15 7134-Recirc. Cooling Рщ-3 Run 1250№(i:SSU) Run 1753 13 (12So ■■

90 IS 7134-Rtcik. Cooling fuap-l

ûfF-SràûR ÜReset (tot Run) Reset (Xu! Run) -

100 17 u 7190-Chiller Circ. Pmp (3) OFF-SmOR Reset (Rot Run) Reset (Xc! R’un) 3 Pumps out of 8 re.].

105 18 tt 7190-Chiller Onit-CH.09 OPP-MW Run 50SU (456.3 tí) Run 500 tí (45S.3 tí!

115 19 it 7190-Chiller Onit-CB.il 0FF-0R§ Run 509 № (456.3 it) Run 500 tí (45S.3 tí) Operates 501 of time, tod 228kR

125 20 tt 3331-8. ft feed Pimp OFF-Ш Ш Run 447 tí (283 tím Run 417 tí (283 tí!

135 21 tt 7512-Instiiir Compressor

1 * -0FP-A0ljjjSTBY-Ç5x Run 220 tí (190 tíj)

Run 320 tí ¡190 tí! Operates 501 of time.

140 22 tt 7512-Instiir Coapressor

1-OPP-AU$STBT-08'

1

Reset (Rot Run) |’i, Reset (Rot Run)

.'¡'И1:'! I jf 4

~1

■VIEiCc-J

ms:1.

2.3.4.5. S. 7.

iced

IssiLve

?■ IOpeating tods are shoo is ршлthesis. vLosses of ISO tí should be added to detenine the sixe of the №. final пшвЬег to be mimed bp MSALDJ. The rated tods shorn are the ootor shaft output. >The elapsed tine is after the К CB closes. iLoad sequence mist he coupleted vithin ISO seconds ftm receijfefif LOCLIY afoul.** indicates loads пЦ one К on to bus in case ofjfrSOt,In case of 2-100t SGs/Odd sequence till be аррИсаЬЩ to bod v uith Mtot rariatfjj&of HCC tods.

s.iDiidir. ншт

Odd. Sif