safety assessment and improvements in indian nuclearsecure site · p.krishna kumar,...

Safety Assessment and Improvements in Indian NuclearImprovements in Indian Nuclear

Power Plants

P.Krishna Kumar, S.Hajela,P.K.Malhotra & S.G.GhadgeNuclear Power Corporation of India Ltd

International Conference on Opportunities and Challenges for Water Cooled Reactors in the 21st Century, 27-30 October 2009

International Conference on Opportunities and Challenges for Water

Cooled Reactors in the 21st Century, 27-30 October 2009

Introduction of NPCILIntroduction of NPCIL

Formed in 1987 for accelerated growth.

Profit making company.

17 reactors in operation of installed capacity 4120 MWe.17 reactors in operation of installed capacity 4120 MWe.

5 Reactors under construction of total capacity 2660 MWe.



4

NPCIL - PROGRESS SO FAR 2000sEXPANSION,

COMMERCIALISATION &ECONOMY OF SCALES

700 MWe & above

FUTUREPROJECTS

1990s2000s

COMMERCIALISATION

220 MWe540 MWe

1980s

1980sSTANDARDISATION

1990sCONSOLIDATION

TAPS-3&4

1970sTECHNOLOGY

DEMONSTRATION

INDIGENISATION

RAPP-3&4 KAIGA-3&4

RAPS-1&2 MAPS-1&2 NAPS-1&2 KAPS-1&2 KGS-1&2 RAPP-5&6

With NPCIL’s tenet to continuously improvey pperformance of country’s nuclear power plantswithout compromising safety; the safetyp g y; yassessment and improvements are primarilydriven by:y

• Evolution of technology

• Feedback from experience both indigenous and• Feedback from experience, both indigenous and international

h fi di• New research findings

• Evolving regulatory requirements



TAPSTAPS--1&21&2Turnkey Project by GETurnkey Project by GE USA 2x160 MWe BWRs2x160 MWe BWRsUnits 1 & 2 : Oct. 69

DesignDesign andand EngineeringEngineering S pportS pport forfor operationoperation pro idedpro idedDesignDesign andand EngineeringEngineering SupportSupport forfor operationoperation providedprovidedindigenously,indigenously, particularlyparticularly sincesince 19741974,, IncludingIncluding FuelFuel redesign,redesign,Physics,Physics, etcetc……..AA majormajor safetysafety analysis,analysis, seismicseismic reanalysisreanalysis andand lifelife extensionextensionstudystudy andand consequentconsequent extensiveextensive refurbishmentrefurbishment carriedcarried outout -- lifelifeextendedextended..



extendedextended..

Life extension and Safety Up gradation of BWR,

TAPS‐1&2

• A comprehensive review of the plant including station-operatingperformance, ageing assessment & management, design basis &safety analysis and structural integrity studies, after 35 years ofoperation using relevant current safety standards and practicesconcluded that the physical condition of the Systems Structures andconcluded that the physical condition of the Systems, Structures andComponents (SSCs) permits continued operation for several moreyears. This indepth review identified certain activities to becompleted for common safe operation. After successful completioncompleted for common safe operation. After successful completionof all the identified activities, the units were restarted in February -2006 after obtaining AERB authorisation for operation of the unitsfor next five years.



Design BasisThe salient recommendations from the safety review include:The salient recommendations from the safety review include:•Up gradation of 3x50 % Emergency Diesel Generators (EDGs) by 3x100% EDGs.•Unit wise segregation of shutdown cooling system and de‐linking it from fuel poolcooling systemcooling system•Segregation of Class‐III, II and I power supplies into two zones with a physicalbarrier and redistribution of supply to redundant loads from separate buses.•The existing shared shutdown cooling system was upgraded to make itThe existing shared shutdown cooling system was upgraded to make itindependent for each unit by installing additional pump and heat exchangers. Thefuel pool cooling system has also been made independent of shutdown coolingsystem.•Rerouting of cable through diverse routes for redundant loads.•Augmentation in fire protection system•Up gradation of electrical power supplies to important loads•Provision of supplementary control room



Retrofitting of Components and Upgradation f ld PHWR ifor old PHWR units

Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS) are theearlier generation Pressurized Heavy Water Reactor (PHWR) in India The design philosophyearlier generation Pressurized Heavy Water Reactor (PHWR) in India. The design philosophyof later PHWRs has undergone continuous improvements. While it is not possible toimplement all the new concepts, certain modifications and retrofitting are incorporated inthese old units to bring their safety standards close to that of new plants.

These include:These include:• Retrofitting of high pressure ECCS,• Incorporation of supplementary control room,• Replacement of thermal shield cooling coils,

S i f / l bl• Segregation of power/control cables,• Replacement of coolant tubes and feeders• Provision of dedicated instrument air supply to safety related valves,• Installation of Flood Diesel Generator in RAPS‐1&2• Installation of Emergency Diesel generator, Fire water pump and air compressors at higher

elevation.



Retrofitting of high pressure ECCSRetrofitting of high pressure ECCS• The existing Emergency Core cooling System (ECCS) in MAPS‐1 consists of low

pressure moderator water injection/recirculation into Primary Heat Transport (PHT)system in case of a loss of coolant accident (LOCA) This system could adequatelysystem in case of a loss of coolant accident (LOCA). This system could adequatelytake care of the large breaks in PHT system. However, for small/medium breaks, alonger time is expected for the PHT system to fall down to a value sufficient toenable the moderator injection. During this period, void formation may take placein the channel leading to high fuel temperature Hence the retrofitting of the ECCSin the channel, leading to high fuel temperature. Hence, the retrofitting of the ECCSsystem has been carried out to mitigate this situation.

• The retrofitted ECCS consists of high‐pressure heavy water (D2O) injection followedb d f d l l d / l lby modified low pressure long term moderator water injection /recirculation at lowpressure. The retrofitting with high pressure heavy water injection and othermodifications are carried out to strengthen the existing ECCS for minimizing thefuel failures during LOCA, covering the entire range of break sizes.



REACTOR BUILDING - UNIT - 2

RIH ROHROH

ECCS BUILDING

LEGEND :EXISTINGMODIFIED

REACTOR BUILDING - UNIT - 1



CCS U G

Provision of Supplementary C l R (SCR)Control Room (SCR)

• Supplementary control room has been provided to ensure safe shutdown of the reactorin case of loss of inhabitability in main control room. It will also help in maintainingreactor in a safe shutdown state ensuring core decay heat removal for extended periodreactor in a safe shutdown state, ensuring core decay heat removal for extended periodof time by monitoring essential plant parameters. Supplementary control panel has thefollowing provisions.

– Tripping of the reactor.– Opening of 2 nos. of the Atmospheric Steam Discharge Valves.(ASDV)p g p g ( )– Monitoring of all essential plant parameters like PHT system pressure, temperature, boiler

pressure, radiation field in boiler room, boiler room pressure and temperature, calandria leveland calandria moderator outlet temperature.

– Lamp indications for boiler level low/high conditions and reactor building isolation damperpositionspositions.

– Recording of neutronic parameters, Log N and Log Rate N– A terminal of plant information system (PIS) to monitor various plant parameters.

• SCR is designed in such a way that it is physically and electrically isolated from maincontrol room Except in the case of neutronic parameters separate sensors are used forcontrol room. Except in the case of neutronic parameters, separate sensors are used formonitoring important process parameters and they are routed separately to SCR.



Replacement of coolant tubes

• In the earlier design of PHWRs used at RAPS‐1&2, MAPS‐1&2, NAPS‐1&2 and KAPS‐1 azirconium alloy (Zircalloy 2) was used for coolant channels. It was considered the best

il bl i l h i i il i b h h iavailable material at that time. However, in pile experience brought out, the requirementof replacement of coolant tubes after 10 to 12 effective full power years in view of themodification of material characteristics especially the reduction in mechanical strength dueto hydriding under radiation during service. The job of Enmass Coolant ChannelReplacement (EMCCR) at RAPS‐2 attempted first time with indigenous technology wasReplacement (EMCCR) at RAPS‐2, attempted first time with indigenous technology, wascompleted successfully in a record time and all the coolant channels were replaced by animproved coolant tube material ‐Zirconium ‐ 2.5% Niobium. This was for the first time in adeveloping country and only the second time in the world that such a highly complextechnical project was completed. The job was completed using indigenously developedremotely handled tools. Similar EMCCR work been accomplished at MAPS 1&2 & NAPS‐1,is underway at NAPS‐2 & KAPS‐1.



Replacement of Feeders in h P i C lthe Primary Coolant system.

• Flow assisted corrosion leading to thinning of the carbon steel feederlb h l f d delbows at the outlet of reactor was noticed in some Canadian PHWRsin 1996. A study and assessment was carried out in‐house to detectany thinning of feeder pipes in Indian PHWRs. However, significantthinning of feeders was not noticed in Indian PHWRs The reducedthinning of feeders was not noticed in Indian PHWRs. The reduceddegradation of feeders in Indian PHWRs was attributed to goodchemical control and better operational practices in Indian reactors.For the earlier PHWRs RAPS‐2 and MAPS‐1&2, which had seen a long, gservice life, a decision was taken to replace the feeders to extendtheir life. Hence at MAPS‐1 the Enmasse Feeder Replacement (EMFR)was carried out along with the EMCCR work. This was for the firsti i h ld h f d l d i PHWR EMFR htime in the world that feeders were replaced in a PHWR. EMFR hasalso been carried out at NAPS‐1 & RAPS‐2.



Installation of Flood Diesel GeneratorInstallation of Flood Diesel Generator• The older PHWR units at RAPS and MAPS had fixedtheir safe grade elevation on certain basis. Withemerging regulatory requirements, it is seen thatlater plants at these locations have hyper gradeelevations. This is an area where no back fitting isgpossible. However to ensure continued power supplyfor essential functions; one emergency dieselgenerator at these stations is located at elevationgcorresponding to the later plants at these sites. Inaddition, in MAPS, one diesel operated fire waterpump and air compressor is also located at higherpump and air compressor is also located at higherelevation.



Accident ManagementAccident Management

• Indian PHWRs have event based EmergencyIndian PHWRs have event based EmergencyOperating Procedures (EOPs) to handle accidentswithin the design basis. These procedures are beingcomplemented with the symptom based guidelines,which are developed to maintain identified ‘safetyf ti ’ it d ti l th h thfunctions’ ‐ monitored continuously through theidentified set of plant parameters (symptoms) andmaintained in the acceptable statemaintained in the acceptable state.



Severe Accident for PHWR Starts withSevere Accident for PHWR Starts with

LOCA + ECCS Failureand progresses withand progresses with

Loss of Moderator Cooling dand

Loss of Calandria Vault Waterin succession



LOCA AND ECCS FAILURE

The moderator system will act as an effective heat sink for this type of situationsof situations.



LOCA + ECCS FAILURE+ Moderator Cooling FailureSevere Accident Involving Core

Damage

If moderator cooling not

g

cooling not available, moderator system

Rupture DiskRupture Disk

Calandria Vault

will pressurize connecting to containment

Calandria Vessel

containment atmosphere via rupture disksrupture disks



LOCA + ECCS FAILURE+ Moderator Cooling Failure

��

The moderator expulsion from

��

expulsion from calandria will uncover upper fuel channels

��

��

ppwhich will rapidly fail ��

��

��



LOCA + ECCS FAILURE+ Moderator Cooling Failure

The Debris will collect at the bottom of Calandria



LOCA + ECCS FAILURE+LOCA + ECCS FAILURE+ Moderator Cooling Failure



LOCA + ECCS FAILURE+ Moderator Cooling Failure + calandria vault water cooling failure

If calandria vault water cooling is lost, The calandria fails &The calandria fails & molten debris falls on the floor of calandria vaultcalandria vault

Core Concrete interactioninteraction



Severe Accident Management for IPHWRs, for which work isinitiated, appropriate intervention in the form of strategies will beid tifi d t i t i / t t l t h l i t it l d iidentified to maintain/protect coolant channel integrity, calandria,calandria vault and containment, which act as ‘barriers’ in severe

accident progression. These Strategies willaccident progression. These Strategies will

• Exploring full design capability of the plant including possibleuse of some systems (both safety and non‐safety ones)y y ybeyond their originally intended function.

• Use of additional temporary system/ad‐hoc arrangements toreturn the plant to a controlled safe state

• Getting support from other units in a multi unit stationprovided safe operation of the other units is notcompromised.

• Identifying instruments that could aid handling of severeaccident situations



Severe Accidents Management Guidelines (SAMGs) are under development for Indian PHWRs. Following strategies p g g

are envisaged

– Inject into PHT system– Maintain calandria heat sinkMaintain calandria heat sink– Maintain calandria vault heat sink– Control Reactor Building conditionControl Reactor Building condition– Reduce Containment Pressure– Control Containment Atmosphere flammability p yand H2

– Mitigate fission product release



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