Training Report Npcil Rapp Rawatbhata

AREPORT ON VOCATIONAL TRAINING

NUCLEAR POWER CORPORATION OF INDIA LTDNUCLEAR POWER CORPORATION OF INDIA LTD..(A Government of India Enterprise)

Rajasthan Atomic Power Station

DURING THE PERIOD

FROM 11th JUNE 2012 TO 10th JULY 2012

SUBMITED TO: Mr. M. M. GUPTA SUBMITTED BY

UMESH KUMAR MEHARB.TECH. (III Year) BRANCH: - ECE

SUBMITTED BY: - UMESH KUMAR MEHAR RAJASTHAN INSTITUTE OF ENGINEERING & TECHNOLOGY,CHITTORGARH

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PREFACE

As we know that an engineer has to serve an industry, for that one must be aware of industrial

environment, their management, problems and the way of working out their solutions at the

industry.

After the completion of the course an engineer must have knowledge of interrelation between

the theory and the practical. For this, one must be familiar with the practical knowledge with

theory aspects.

To aware with practical knowledge the engineering courses provides a six weeks industrial

training where we get the opportunity to get theory applying for running the various process and

production in the industry.

I have been lucky enough to get a chance for undergoing this training at RAJASTHAN

ATOMIC POWER STATION. It is a constituent of board of NPCIL. This report has been

prepared on the basis of knowledge acquired by me during my training period of 30 days at the

plant.


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ACKNOWLEDGEMENT

It was highly educative and interactive to take training at

RAJASTHAN ATOMIC POWER STATION. As technical knowledge

is incomplete without practical knowledge, I couldn’t find any place

better than this to update myself.

I am very much thankful to the Site director Mr. C.P. Jhamb &Training

superintendent Mr. D. Chanda for allowing me for the industrial training

at RAPS. Thanks to Mr. A.P. Jain for their guidance during my project.

I also take the opportunity to thanks Nuclear training Centre for

providing lecture on overview of the plant and providing me Orange

qualification.


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INTRODUCTION

India's Nuclear power developments are under the purview of the

Nuclear Power Corporation of India, a government-owned entity under

the Department of Atomic Energy India. The corporation is responsible

for designing, constructing, and operating nuclear-power plants. In

1995 there were nine operational plants with a potential total capacity

of 1,800 megawatts, about 3 percent of India's total power generation.

There are two units each in Tarapur, north of Bombay in Maharashtra;

in Rawatbhata in Rajasthan; in Kalpakkam near Madras in Tamil Nadu;

and in Narora in Uttar Pradesh; and one unit in Kakrapur in

southeastern Gujarat. However, of the nine plants, all have been faced

with safety problems that have shut down reactors for periods ranging

from months to years. The Rajasthan Atomic Power Station in

Rawatbhata, India was closed indefinitely, as of February 1995.

Moreover, environmental problems, caused by radiation leaks, have

cropped up in communities near Rawatbhata. Other plants operate at

only a fraction of their capacity, and some foreign experts consider

them the most inefficient nuclear-power plants in the world.


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MISSIONTo develop nuclear power technology and

produce in a self-reliant manner nuclear

power as a safe, environmentally benign and

an economically viable source of electrical

energy to meet the growing electricity needs

of the country

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VISIONNPCIL has its vision to have an

installed nuclear power capacity

of 20,000 MW(e) by the year

2020. This capacity could be

achieved by the development of

more 220 MW(e) & 550 MW(e)

units of Pressurized heavy water

reactors, importing light water

reactors and by the introduction

of fast breeder reactors.

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Prime Minister

DAE Atomic Energy Commission

NPCIL

Atomic energy Regulatory board

TAPS 1&2

India Rare Earth

TAPP 3&4

BARC

RAPS 1& 2

Heavy Water Board

RAPS 3 & 4

ECIL

RAPP 5 & 6

UCIL

MAPS

Nuclear fuel complex

NAPS

Indra Gandhi center for advance research

KAPS

KAIGA PS 1& 2

Center for advance technology

KAIGA Proj. 3& 4

KKPS

RAPS LOCATION AND SITE CONDITIONS

RAPS is located on the eastern bank of Rana Pratap Sagar lake (R.P.S) upstream of the R.P.S dam across the chambal river at an elevation of 388 mt. above mean sea level with a latitude of 24053’ north and a longitude of 76036’ east. The plant site is about 64 KM from Kota city. The place has an average rainfall of 825mm as per records. The maximum wind velocity records so far is 129 km/hr at 120 m. the most predominant wind direction is at 7.90m and 120m heights is North of south west and west of south west respectively.The site has no population with in its vicinity of radius of 5km. It however does have a population of about 58 thousand distributed in the radius of 15 Km. the only nearby major industry is HEAVY WATER PLANT (H.W.P).

NUCLEAR ENERGY: Mass defect converted into energy through nuclear reaction. Two processes produce this:1) Nuclear fission.2) Nuclear fusion.A neutron it splits into two big parts hits when a heavy nucleus likes that of uranium – 235 & in addition 2 or 3 neutrons are released. However, the mass of the parts is slightly less than the mass of the uranium nucleus. The mass that is destroyed is converted into energy (200Mev/ fission). This process is called nuclear fission reaction.It is much more likely if neutrons are slow, in a reactor, some of the neutrons produced are absorbed so that for every neutron causing fission, only one is left. This neutron in turn collides with another U235 nucleus & causes fission. A chain reaction is thus set up. Also, the neutrons have to be slowed down. The fuel in a nuclear reactor consists of Uranium that may be natural or enriched in which proportion of U235 is increased. Either light water (for enriched uranium) or heavy water (for natural uranium) may be used as a moderator, for slowing down the neutrons. The energy released is absorbed by the water (either light or heavy). This coolant in turn transfers its energy to the light water. Ultimately water is turned into steam at high pressure that is used to derive turbines as in any conventional power plant. India has six Nuclear Power Plants;


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At Tarapur in Maharastra. At Rawatbhata near Kota in Rajasthan At Kalpakkam near Madras in Tamil Nadu. At Narora in Uttar Pradesh At Kakarpara near Surat in Gujarat At Kaiga near Karwar in Karnataka.The reactors at Tarapur use enriched uranium as fuel & light water as moderator and coolant, all others use uranium and heavy water. Nuclear Power Plant under construction is two units of 500 MW at Tarapur and two similar units at Rawatbhata near Kota. Nuclear fission has become commercially viable and is being exploited in several countries.

SOME IMPORTANT NUCLEAR REACTIONS:

1) 92U238+0n1-----92U239+r------93Np239-------94Pu239

Typical fission reaction: 2) 92U235+0n1------38Sr90+54Xe144+20n1+r+200MeVReactor poisoning reaction: 3) 52Te 135 ----53I135-----54Xe135-----55Cs135---56Ba135

(Stable)We know that about 200MeV of energy is released during per fission.This energy is divided in the following way:1) K.E. of the fission fragments: 167MeV.2) K.E. of neutrons: 5MeV.3) Energy of gamma released at fission: 5MeV. 4) Energy of gamma rays released on n–capture: 10MeV.5) Gamma decay energy: 7MeV.6) Beta – decay energy: 5MeV. ---------------------------- TOTAL =199MeV. ----------------------------


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THREE STAGES OF INDIAN NUCLEAR POWER PROGRAMME:

1) INTRODUCTION:India figured in the nuclear power map of the world in 1969, when two boiling water reactors (BWRS) were commissioned at Tarapur (TAPS-1&2) these reactors were built on the turnkey basis .The main objective of setting these units was, largely to prove the techno-economic viability of nuclear power. The nuclear power programme formulated embarked on the three-stage nuclear power programme, linking the fuel cycle of pressurized heavy water reactor (PHWR) & Fast breeder reactors (FBR) for judicious utilization of our reserves of Uranium & Thorium. The emphasis of the programme is self –reliance and thorium utilization as a long -term objective.

The three stages of our Nuclear power programme are:

1) STAGE I:- This stage envisages construction of natural Uranium, Heavy water moderator & cooled pressurized heavy water reactors (PHWR). Spent fuel from these reactors is reprocessed to obtain plutonium.

2) STAGE II: - This stage envisages on the construction of Fast breeder reactors (FBR) fuelled by plutonium & depleted U produced in stage I. These reactors would also breed U233 from thorium.

3) STAGE III = This stage would comprise power reactors using U233- Thorium as fuel, which is used as a blanket in these type of reactors.

The PHWR was chosen due to the following:

1) It uses natural uranium as fuel. Use of natural uranium available in India, helps cut heavy investments on enrichments, as uranium enrichment is capital intensive.

2) Uranium requirement is the lowest & plutonium production is the highest.


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3) The infrastructure available in the country is suitable for undertaking manufacture of the equipment.The short –term goal of the programme was to complement the generation of electricity at locations away from coalmines. The long-term policy is based on recycling nuclear fuel and harnessing the available Thorium resources to meet country’s long-term energy demand and security. As a part of PHWR Programme (STAGE I) second nuclear power plant was taken up as a joint Indo-Canadian venture this plant was built at Rawatbhata (Rajasthan) two units laid a milestone in the history of India all the components were taken up in India and the import content reduced considerably. Moreover, Canadians withdrew in 1974; Indian engineers did balance design & commissioning of the Unit 2.

2) CHALLENGES FACED:

The industry was new to the manufacturing techniques & stringent quality requirements of the nuclear components like calandria, end shield, steam generators, fuelling machine, and heavy water pumps. The requirement of convectional power plant equipment was of much larger capacity than those being manufactured in the country. To achieve self –sufficiency in this field in the long run, the department of atomic energy established extensive research & development facilities covering diverse areas for supporting technology absorption. Facilities, from prospecting to mining to fabrication of fuel & zirconium alloy components, for manufacture of precision reactor components & production of heavy water were also set up. Supply of equipments of international nuclear standard was also a problem so momentous efforts were put into development of such manufacturing industries. Extensive R&D set up were established for metallurgical studies of both fresh as well as radioactive material, non –destructive testing, environmental & seismic qualification of safety analysis, preparation &development of validation of computer codes, etc.


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Technologies for inspection of the reactor components, repair &replacement using robotics & life extension programme of the operating reactors, have also been successfully developed. To summaries, the concerted efforts put in by DAE, its constituent units & NPCIL, together with Indian industries & institutions have led to development & full capabilities to design, manufacturing of equipment, construction, operation & maintenance of nuclear power plant. Today India is amongst the select band of few countries of the world that have developed such capabilities.

3. Status of nuclear power generation & future plans: The nuclear power programme in India up to year 2020 is based on installation of a series of 235 MWe &500Mwe pressurized heavy water reactor (PHWR) UNITS, 1000MWe light water reactors (LWR) UNITS & fast breeder reactors (FBR) units. NPCIL plans to contribute about 10% of the total additional needs of power of about 10000MWe per year i.e. 1000 MWe per year in the coming two five –year plans. The total installed capacity of nuclear generation would increase to more than 20000 MWe in year 2020 from the present level of 2720 MWe. The basic design of the 220/500MWe units in similar; however, a number of significant design changes have been made progressively from the first unit at Rajasthan to the 500 MWe units. These design changes have been made from the consideration of currently prevailing safety criteria, seismicity, improve availability requirement of in- service inspection, ease of maintenance etc., as appropriate to the conditions in India. DESCRIPTION OF STANDARD INDIAN PHWR:

1) LAYOUT: The nuclear power stations in India are generally planed as two units modules, sharing common facilities such as service building, spent fuel storage bay & other auxiliaries like heavy water upgrading, waste management facilities etc. . Separate safety related systems & components are however provided for each unit. Such an arrangement retains independence for safe operation of each unit & simultaneously permits optimum use of space, finance & construction time. The lay out for a typical


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220MWe station as given in figure 1, shows two reactor building, active service building including spent fuel bay, safety related electrical & control buildings and the two turbine buildings. Orienting turbine building radial to the reactor building provides protection from the effect of turbine missiles. Other safety related building s &structures are also located as not to fall in the trajectory of missiles generated from the turbine. The buildings and structures have also been physically separated on the basis of their seismic classification.Sectional views of the reactor building are shown in figure 2 depicting general layout inside the reactor building.

2) REACTOR: In concept, the Indian pressurized heavy water reactor is a pressure tube type reactor using heavy water moderator, heavy water coolant & natural uranium dioxide fuel. The reactor as shown in the cut away view in figure 3 consists primarily of calandria a horizontal cylindrical vessel. It is penetrated by a large number of zircaloy pressure tubes (306 for 235MWe reactor), arranged in a square lattice. These pressure tubes, also refer as coolant channels, contain the fuel & hot high – pressure heavy water coolant. The pressure tubes are attached to the alloy steel and fitting assemblies at either end by special role expended joints. A typical pressure tube assembly is shown figure 4 .End – shields are the integral parts of the calandria and are provided at each end of the calandria to attenuate the radiation emerging from the reactor, permitting access to the fuelling machine vaults when the reactor is shutdown. The end fittings are supported in the end shield lattice tubes through bearing, which permit their sliding. The calandria is housed in a concrete vault, which is lined with zinc metallised carbon steel & filled with chemically treated demineralised light water for shielding purposes. The end shields are supported in openings vault wall, and form part of the vault enclosure at these openings. Removable shield plugs fitted in the end fittings provide axial shielding to individual coolant channels.


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3) REACTIVITY CONTROL MECHANISMS:

Due to the use of natural uranium fuel & on-load refueling, the PHWR’s do not need a large excess reactivity. Correspondingly the devices required for control of reactivity in the core need not have large reactivity worth’s. Standard reactors designs are provided with four systems for reactivity control, viz.1) Regulating rods 2) Shim rods 3) Adjuster rods for xenon override 4) Natural boron addition in the moderator to compensate for the excess reactivity in a fresh core &for absence of xenon after a long shutdown.The reactivity control devices are installed in the low- pressure moderator region & so they are not subjected to potentially severe hydraulic & thermal forces in the event of postulated accidents. Furthermore, the relatively spacious core lattice of PHWR allows sufficient locations to obtain complete separation between control & protective functions. The regulating systems are thus fully independent with its own power supplies, instrumentations & triplicated control channels. Cobalt & stainless steel absorber elements have been utilized in the reactivity control mechanisms. For 220MWe standardized design, two diverse, fast acting & independent shutdown systems have been adopted. This feature provides a high degree of assurance that plant transients requiring prompt shutdown of the reactor will be terminated safely. The primary shutdown system consists of 14 mechanical shut off rods of cadmium sandwiched in stainless steel &makes the reactor sub- critical in less than 2 secs. Fail-safe features like gravity fall &spring assistance have been incorporated in design if mechanical shut off rods. The second shutdown system, which is also fast acting, comprises 12 liquid poison tubes, which are filled with lithium penta borate solution under helium pressure. The trip signal actuates a combination of fast acting valves and causes poison to be injected simultaneously in 12 interstitial liquid poison tubes of calandria.


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4) FUEL DESIGN:

Fuel assemblies in the reactor are short length (half metre long) fuel bundles. Twelve of such bundles are located in each fuel channel. The basic fuel material is in the form of natural uranium dioxide a pellet, sheathed &sealed in thin zircaloy tubes. Welding them to end plates to form fuel bundles assembles these tubes. Figure 5 shows the 19- element fuel bundle being used in 220 MWe PHWRs. 5) FUEL HANDLING:

On –power fuelling is a feature of all PHWRs, which have very low excess reactivity. In this type of reactor, refueling to compensate for fuel depletion & for over all flux shaping to give optimum power distribution is carried out with the help of 2 fueling machines, which work in conjunction with each other on the opposite ends of a channel. One of the machines is used to fuel the channel while the other one accepts the spent fuel bundles. In addition, the fueling machines facilitate removal of failed fuel bundles. Each fuelling machine is mounted on a bridge & column assembly. Various mechanisms provided along tri- directional movement (X, Y&Z direction) of fueling machine head and make it possible to align it accurately with respect to channels. Various mechanisms have been provided which enables clamping of fueling machine head to the end fitting, opening & closing of the respective seal plugs, shield plugs &perform various fuelling operations i.e. receiving new fuel in the magazine from fuel transfer system, sending spent fuel from magazine to shuttle transfer station, from shuttle transfer station to inspection bay & from inspection bay to spent fuel storage bay.

6) PRIMARY HEAT TRANSPORT (PHT) SYSTEM:

The system, which circulates pressure coolant through the fuel channels to remove the heat generated in fuel, is referred as Primary Heat Transport System. The major components of this system are the reactor fuel channels, feeders, two reactor inlet headers, two reactor outlet headers, four pumps &interconnecting pipes & valves. The headers steam generators & pumps are


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located above the reactor and are arranged in two symmetrical banks at either end of the reactor. The headers are connected to fuel channels through individual feeder pipes. Figure 6 depicts schematically the relative layout of major equipment in one bank of the PHT system .the coolant circulation is mentioned at all times during reactor operation, shutdown & maintenance.

7) MODERATOR SYSTEM:

The heavy water moderator is circulated through the calandria by aid of a low temperature & low – pressure moderator system. This system circulates the moderator through two heat exchangers, which remove heat dissipated by high – energy neutrons during the process of moderation. The cooled moderator is returned to the calandria via. Moderator inlet nozzles. The high chemical purity and low radioactivity level of the moderator are maintained through moderator purification system. The purification system consists of stainless steel Ion – Exchange Hoppers, eight numbers in 220MWe contains nuclear grade, mixed Ion - Exchange resin (80% anion & 20% cation resins) .the purification system is also utilized for removable of chemical shim, boron to effect start –up of reactor Helium is used as a cover – gas over the heavy water in calandria. The concentration deuterium in this cover- gas is control led by circulating it using a sealed blower and passing through the recombination containing catalyst Alumina – coated with 0.3% Palladium.

7) FUEL:

The use of natural uranium dioxide fuel with its low content of fissile material (0.72% U-235) precludes the possibility of a reactivity accident during fuel handling or storage. Also, in the core there would no significant increase in the reactivity, in the ever of any mishaps causing redistribution of the fuel by lattice distortion or otherwise.The thermal characteristics namely the low thermal conductivity and high specific heat oh UO2 permit almost all the heat generated in a fast power transient to be initially absorbed in the fuel. Furthermore, high melting point of UO2 permits several full power


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seconds of heat to be safely absorbed above that contained at normal power.Most of the fission products remain bound in the UO2 matrix and may get released slowly only at temperatures considerably higher than the normal operating temperatures. Also on the account of the uranium dioxide being chemically inert to the water coolant medium, the defected fuel releases limited amount of radioactivity to the primary coolant system.The use of 12 short length fuel bundles per channels in a PHWR, rather than full – length elements covering the whole length of the core, subdivides the escapable radioactive facility in PHWR has also the singular advantage of allowing the defected fuel to be replaced by fresh fuel at any time. The thin Zircalloy – 2/4 cladding used in fuel elements is designed to collapse under coolant pressure on to the fuel pellets. This feature permits high pellet - clad gap conductance resulting in lower fuel temperatures & consequently lower fission gas release from the UO2 matrix into pellet – clad gap.

REACTOR AUXILIARIES

END SHIELD COOLING SYSTEM

There are two End Shields provided at both the ends of calandria performing the following functions.(i) Providing supports for calandria tubes and pressure tubes.(ii) Provides radiation and thermal shielding for fuelling machine vaults so that the fuelling machine vaults can be accessible during shutdown.Heat is removed from the end shields to moderator and calandria vault water. However the bulk of the heat is removed by End shield cooling system.

The basic requirements of the end shield cooling system are:

(i)To maintain calaridria side tube sheet (CSTS) of end shield at an averagetemperature of 67deg centigrade.


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(ii)To maintain temperature difference between various parts of end shieldwithin permissible limits.(iii) To avoid stagnant pockets of coolant, in end shield, which could causecorrosion problems.(iv)To avoid overheating and hot spots which could lead to damage of endshield.(v)To provide venting of end shield for uniform shielding in accessible andS/D accessible areas.

The End Shield Cooling System is a closed loop system Consisting of end shields, circulating pumps, and heat exchangers. An auxiliary loop exists for the control of water chemistry.There are two end shields where the heat is generated due to radiation and conduction from other reactors component i.e. End fittings, Feeders, convection andradiation across insulation gaps. (Almost 50% of the

heat load is from PHT). A total of 1.4 MW of heat loadexists for each end shield. This heat is removed by

__ circulation of demineralised water through the EndShields. The End Shields consist of two compartments called front and rear compartments. DM Water (900 LPM)enters the front compartment (the compartment facing the calandria) from five inlets at the top. Front Compartment is further divided into five separate columns. DM Water passes through these columns at a velocity of 37.7 cm/sec and flows into the annulus space between the outer and inner shells of End shield.

CALANDRIA VAULT COOLING SYSTEM

In RAPS calandria vault (the space between the calandria and steel lined structural wall) is full of demineralised (DM) water. DM water filled calandria vault provides radiation, biological and thermal shielding, and also acts as heat sink in case of serious contingency. Filling of calandria vault with DM water eliminated Argon-41 activity of earlier Indian PHWRs which had air filled


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calandria vaults (RAPS 1&2 AND MAPS). This drastically cuts the exposure of public in the vicinity of Indian Nuclear Power Plants.

The dimensions of the calandria vault are such that a minimum water thickness of 1.35 meters is ensured between the calandria and concrete vault.This ensures adequate shielding.

FUNCTIONS OF THE CALANDRIA VAULT COOLING SYSTEMi)To remove heat generated in vault water.

ii)To provide thermal shielding and biological shield under all condition.

iii) To maintain uniform temperature in the vault structure below permissible limit under all condition.

iv) Provide an environment compatible with the material used for components within vault.

Heat appearing in calandria vault water is removed by a closed loop cooling system. Water at 42.5deg cen. is distributed through perforated header laid out in the bottom of the vault and warm water at 46.2deg cen. leaves the vault through header at the top.

VAPOUR SUPPRESSION SYSTEM

Large pooi of water (2200M3, 2.4m deep) at the basement of the reactor building is provided to limit peak pressure inside volume Vi during LOCA (Loss of coolant accident) or MSLB(Main steam line break) by condensinghigh enthalpy steam. Volume Vi is connected to the suppression pool via an annular space between the RB structure wall and inner containment wall.The suppression pool is provided with a re circulation system to protect against corrosion and biological growth.


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ANNULUS GAS MONITORING SYSTEM

The annulus gas monitoring system of RAPP 3&4 provides a means of monitoring the leakage (if any) of heavy water either from PHT or from moderator system due to failure of coolant tube calandria tube or rolled joints. It is a closed loop recirculating system which maintains flow of C02 gas through the annulus gap between coolant tithe and calandria tithe. Apart from leak detection, the annulus gas also acts as a thermal barrier, separating the hot high pressure coolant tubes and the comparatively cooler low pressure calandria tubes. By reducing heat transfer between coolant tube and calandria tube, heat removal requirements from moderator system are minimized as well as the reduction in loss of heat from PHT system. In addition, the annulus gas minimizes corrosion and hydrides formation in the coolant tubes or in the garter spring spacers by providing a dry 02 doped gas atmosphere in the annulus.

LIQUID POISON INJECTION SYSTEM

For prolonged shutdown of reactor (1) for maintenance jobs or (ii) when reactor has tripped on reactivity transient which do not permit restart of reactor within poison override time, LPIS is actuated so that sub criticality margin is maintained under all conditions. LPIS adds a bulk amount of liquid poison directly to the moderator to keep the reactor under shutdown state for prolonged duration. This is an independent process system and is the replacement of (i) ALPAS bulk addition mode (at NAPP and KAPP) which required moderator circulation and (ii) gravity addition of boron (GRAB)The LPIS works on pneumatic pressurization of boron solution by helium. The system consists of poison tank

and helium tank. When a command for poison addition is received the pressure balance valves and siphon break valves close and injection valves open. This causes the pressurisation of poison tank by helium stored in helium tank. This in turn causes injection of


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boron poison directly into the moderator through two nozzles in calandria at 75%FT level

D2O EVAPORATION AND CLE~AN UP SYSTEM

D20 evaporation and clean up system purifies downgraded heavy water to a level which is not harmful to heavy water upgrading system by removing all the impurities. The heavy water collected from various leakages and spills contains a number of impurities which normally arise from— Surf ace from which D20 is collected. Corrosion products produced inside the reactor D20 system.Products resulting from radiolytic process. Organic material from ion exchange resin dueteration and breakdown.D20 evaporation and cleanup system is designed to clean the downgraded heavy water chemically so that it can be fed to upgrading plant. Cleanup system comprises of oil water separation stage, filtration stage and ion exchange stage.

HEAVY WATER UPGRADING SYSTEM

Heavy water is used as moderator and primary heat transport fluids in PHWRs. Heavy water is highly hygroscopic. Hence it leaks from the system, it gets downgraded on exposure to atmosphere. Such leaked heavy water collected from various points in the reactor is to be upgraded before use in reactor, since the isotopic purity required for moderator heavy water is as maximum as achievable.

FIRE FIGHTING SYSTEM

Fire protection system in a nuclear power plant is meant To prevent damage to various equipment or system due to fire.To ensure decay heat removal of the reactor. To minimize the release of radioactivity to environment in the event of a fire.To provide backup PW cooling to various systems. To ensure personnel spray supply.Fire protection system consists of fire fighting water system, carbon dioxide fire protection system and portable fire protection system.


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FIRE WATER SYSTEM

Fire water system comprises of constantly pressurized fire hydrant system and sprinkler system. Automatic sprinklers have been provided for oil filled transformers and non-automatic sprinklers are provided for oil systems, cable vaults and cable tunnels. Hydrant system covers the whole plant for outdoor and indoor supply of firewater. Water for both hydrant and sprinkler system is supplied by the firewater pumps from the sumps located in the cooling water pump house(CWPH).

ACTIVE PROCESS WATER SYSTEM

Active process water system provides direct means of heat transport from equipment and heat exchangers in the primary heat transport system, moderator system and reactor auxiliary system to ultimate heat sink during all operational stages of the plant and accident condition like LOCA. Thus it forms the secondary part in the ultimate heat removal system. It is a safety-related system. Reliability and continuous heat removal is achieved by designing the system for SSE/OBE by providing redundancy in rotating equipment, Class III power supply to all safety related electric motor driven equipment and backup supply from fire water system to meet static component failure This system is potentially active since there is a possibility of leakage of active primary fluid to this system through various heat exchangers.

RB VENTILATION SYSTEM

RB is designed as a double containment structure in order to prevent ground level release during accident conditions. Primary containment houses all equipments and piping of nuclear systems. Secondary containment envelops the primary containment with an annular radial gap of 2 meters.PC is divided into two volumes. Vi containing the systems having high enthalpy fluids comprising of F/M vaults, pump room, dome region and includes FMSA when they are in contact with F/M


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vaults. These areas are not accessible during normal plant operation. No ventilation is provided for this volume but closed loop heavy water vapour recovery system is provided to recover D20 that escapes from high enthalpy systems. The remaining area constitutes volume V2. Volume V2 is separated from Vi by a leak tight barrier and pressure suppression pool. The volume Vi is maintained at negative pressure with respect to V2 by maintaining continuously a small purge to the stack. Volume V2 is normally accessible except moderator room, FMSA and DN monitoring rooms.

HEAVY WATER VAPOUR RECOVERY SYSTEM

Heavy water vapour arising out of spills/leakages from primary heat transport, moderator and fuelling machine circuits is recovered from building atmosphere by adsorption on molecular sieve beds. Vapour recovery system is an important feature of the station heavy water management schemes. Following are the criteria for design and operation of vapour recovery system— To effect economy in reactor operating costs by efficient recovery of heavy water that escapes into the building atmosphere.To minimise heavy water loss and tritium loss and tritium release through stack.To minimise tritium activity levels in various areas of the reactor building.

To keep the volume Vi area under negative pressure with respect to volume V2 area for preventing the spread of activity from volume Vi to volume V2.

CALANDRIA

The calandria is horizontal vessel housed in a rectangular calandria vault. The calandria is a single walled austenitic stainless steel vessel. The main shell is stepped down in diameter at each end and


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site welded to their cylindrical extensions of the end shields on each side of the reactor.

END SHIELD

The end shields are cylindrical boxes whose extensions are welded to the calandria side tube sheet at the calandria end and fueling machine side tube sheets at fueling machine end of the end shield during shop fabrication. The box is pierced by 306 lattice tubes arranged on 228.6mm square pitch. The space inside the end shield is divided into two compartments by a 38mm thick baffle plate and fueling machine side tube sheet is filled with 10mm dia spherical mild steel balls and light water in the 57:43 ratio.

CALANDRIA TUBE

The calandria tubes are manufactured from Zircalloy2 strip that is cylindrically formed and seam welded. The seams are then leveled by rolling. The primary functions of the calandria tubes in a reactor system are-

1.To separate the relatively cold moderator from hot coolant tubes to minimize heat losses.2.To support the horizontal coolant tubes (through garter springs) and prevent the excessive sag caused by creep. To act as containment vessel for the contents of the channel in the unlikely but postulated instance of a pressure (coolant) tube rupture accident.

COOLANT TUBE

Coolant tube is the most important structural component inside the reactor core. Coolant tubes are manufactured from Zr-Nb. Each end of the coolant tube is joined to a special type 403SS end


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fitting. Such 306 Nos. of parallel coolant tubes are placed horizontally inside the reactor core at the square lattice distance of 228. 6mm.

END FITTINGS

The end fittings on either end of the reactor identical and connected at the ends.

GARTER SPRING SPACERS

Four numbers of garter spring for each coolant channel and located in the annulus space between coolant and calandria tubes.

SEAL PLUG

The function of the sealing plug is to close the ends of the coolant assemblies and prevent the escape of heavy water from the end fittings. During fuel changes it is necessary to remove these plugs.

SHIELD PLUG

The shield plug which normally resides in the end fitting serves the three functions of providing — Radiation shielding at the ends of the coolant tubes Means of locating the fuel in the fuel channel and stopping the fuel column from following the seal plugs when they are withdrawn during fuel changes. The turbine is of the horizontal tandem compound, reheating, impulse type, running at 3000 rpm, with special provision for extraction of moisture. The turbine has a maximum continuous and economic rating of 220 MW, The turbine comprises of one HP cylinder and two double flow LPCylinders thus providing 4 LP flow in parallel. Thetas are five impulse stages in the HP cylinder and six stages foreach of the LP cylinders. The turbine cylinders and generator is solidly coupled together in line, with a single thrust bearing on HP shaft between No, 2 bearing (HP rotor bearing) and the HP~LP. Coupling each rotor is supported in two main bearing. A solid forged steel rotor is provided in the HP cylinder whilst the LP rotor have shrunk and keyed on discs, The nozzle plates of the HP


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cylinder are welded assemblies incorporating machined nozzle segments, The LP diaphragms are cast iron with cast-in nozzle division plates. Steams packed labyrinth glands are provided for each cylinder, Live steam at a pressure of 580 psig and temp 482.60F (saturated) is supplied to the HP cylinder of the turbine through two separately anchored steam chests each containing a steam strainer a combined stop and emergency valve and two throttle (or governing) valves, The chests are connected to the HP cylinder through loop at allow axial movement of this Cylinder, and ensure that no excessive thrust loads from the piping are transmitted to the HP cylinder. Extraction steam is taken for feed heating purpose before stages 4&5 and at the exhaust of the H.P cylinder after expansion is led two moisture separators in parallel which reduce the moisture content of the steam before it is reheating is two live steam reheaters, The steam from the reheaters Having a pressure of 47.5 psig and temperature of 43OoF passes through governor operated butterfly interceptor valves before entering the two double flow LP cylinders. An interceptor’s valve is provided in the line from each reheater to the LP inlets, The LP cylinder of turbine is of four flow design: each two flow LP has a central admission belt with outward direction of steam flows. Steam is supplied to the two flow provision via the separator and reheater in each HP cylinder .the exhausts from the LP bleeding combines into a condenser, which is maintained at vacuum 27.5” hg.

Steam is extracted from double flow L1~ cylinder before stages 24 and b for feed heating and before stage 6 for a moisture extractor, The over all length of the Turbine generators 100 and the outside diameter of the last row of blade is 100”. A data logger monitors all turbine and ancillaries parameter.

STEAM CYCLE:

Steam for the turbine through two steam lines or header to the two stem lines or header to the two combined stop and emergency valves. A10” balance line connected line connect the header the C.S.E. valves. During normal operation the C~S.E. valves are fully


27

open to permit steam flow to inlet steam chest and then to the two governor valves. Governor valve position controls turbine speed and load and thus are made responsive to the governor valves (two on each bank) are connected by means of balance lines and the steam passes to the H.P. inlet nozzles, trough the H.P. cylinder. The to governor valves in each steam chest are in parallel i.e., there is common inlet and outlet manifold for both of the two governor valves in a steam chest. Also lines from each steam chest joint and go to both top and bottom of H.P, cylinder This arrangement in conjunction with the 10” balance line ensures uniform steam take off from each of the 8 boiler and uniform distribution to each portion of H~P. cylinder, After expansion, the steam leaves the H~P. cylinder and passes through the separator, reheater, LP emergency stops valves and interceptor valves, before entering LP cylinders. Also to relief valves are installed after each of the two re heaters. In the event of governor or interceptor valve malfunction the relief valve will open and vent the steam to atmosphere preventing over pressurizing the separator or reheaters,

The interceptor valves remain full open normal operation and admit the steam to LP cylinder from where it is exhausted to the condenser, not all of the steam to the LP cylinders from where it is exhausted to condenser. Not all of the steam admitted to turbine by the governor valves is expanded through the turbine and exhausted to main condenser At different points on the turbine, steam is bled off or Extracted and passed to feed water heat exchangers. Heating the feed water by extraction steam has two beneficial results; one is an increase in the heat cycle efficiency and the minimum permissible inlet feed temp. to boiler is 24OoF, The RAPS turbine has six extractions feed heaters three including deaerators heatersFed from the H.P, cylinder, and are called lob pressure heaters if they are in the feed line before the boiler feed pumps and are called high pressure if they are in the feed line after the pumps. Five of the extraction lines have spring closed check valves. These check valves close on a turbine trip to prevent entrained stream in the extraction lines and heaters from backing up into the turbine and causing it to over speed. Entrained steam from the (one, remaining extraction line and heater was calculated to give only a small increase so check valves were omitted


28

CONDENSING SYSTEM: GENERAL.

The circulating water in the condenser condenses the exhaust steam from LP Turbines. The condensate is recycled through boilers. The air gases are removed from the condensate by the air ejectors. The condensing system is provided to supply condensate from the deaerators under all condition of operation. The maximum flow of condensate to deaerators at 100% turbine load is i 900.000lbs/hr, the design temp. are 91 5oF in the condenser hot well and 245oF at the deaerators inlet

DESCRIPTION

Condensate system comprises of main condenser two 100% capacity condensate extraction pumps and two 21/4% duty emergency pump, 2 moisture Extractor, gland steam and high level reserve feed water tank with their associate fittings, pipelines and instrumentationThe condensate extraction pumps take suction from the condenser hot well and discharge through the moisture extractors, drain cooler and LP heaters, the condensate flow is controlled by the control valve part of the Condensate front the condensate pump discharge header flow changes the gland steam condenser and air ejectors and returns to main condensate line before it enters the moisture extractors. The flow in this line is controlled by means of regulating control valve which maintains a fixed differential pressure across the gland steam condenser having been designed to have the same pressure deferential tar its design flow as the air ejector. A condensate recirculation line back to the condenser is provided. This take off is located downstream of the condenser. The condensate pumps also supply boiler feed pump gland seal water and water f or the turbine spray cooling. One 21/2% capacity auxiliary condensateExtraction pump takes water from the condenser hot well and discharge into the same system as the 100% pumps.


29

SAFETY DESIGN PRINCIPALS

It has been ensured that systems, components & structures having a bearing on reactor safety are designed to meet stringent performance & reliability requirements. These requirements are met by adopting the following design principles:a) The quality requirements for design, fabrication, construction & inspection for these systems are of the high order, commensurate with their importance to safety.b) The safety related equipment inside the containment building is designed to perform its function even under the elevated pressure & temperature &steam environment conditions expected in the event of postulated loss of coolant accidents (LOCA).c) Physical & functional separation is assured between process systems & safety systems.d) Adequate redundancy is provided in systems such that the minimum safety functions can be performed even in the event of single active components in the system.e) To minimize the probability of unsafe failures f) Provisions are incorporated to ensure that active components in the safety systems are testable periodically.g) All the supplies /services (electric, compressed air or water) to these systems, necessary for the performance of their safety functions are assured & ‘safety grade’ sources.

SAFETY & SEISMIC CLASSIFICATION OF SYSTEMS:

SAFETY CALSSIFICATION:In the design of Indian PHWRs, it is required to grade various systems, equipment & structures in their importance to safety & reliability. The safety gradation consists of four different safety classes depending upon the nature of safety functions to be performed by the various items of the plant.

SAFETY CLASS I: It is the highest safety class & includes equipment & structures needed to accomplish safety functions necessary to prevent release of substantial core fission product


30

inventory. This includes reactor shutdown systems & primary heat transport system.

SAFETY CLASS II: Includes equipment, which performs those safety functions, which become necessary to mitigate the consequences of an accident involving release of substantial core fission product inventory from fuel. This class also includes those items, which are required to prevent escalation of anticipated operational occurrences to accident conditions. Boiler feed water & steam system, emergency core cooling system, reactivity control provisions & reactor containment building are included in this class.

SAFETY CLASS III: Includes systems that perform functions, which are needed to support the safety functions of safety class II & I. Also, it includes systems & functions required to control the release of radioactivity from sources located outside the reactor building. Process water-cooling system include induced draft cooling towers, air supply system, shield cooling system primary coolant purification ion exchange columns & filters etc. are included in this category.

SAFETY CLASS IV: Includes those items & systems, which do not fall within the above classes but are required to limit the discharge of radioactive material & airborne radioactivity below the prescribed limits .D2O upgrading, waste management, dueteration &service building ventilation systems are classified as class IV safety systems.


31


32

1.0 DESIGN DESCRIPTION:

COIS is a data acquisition and display equipment for providing the

operator with process alarm messages, status, trend curves, history

displays and printouts of groups of process variables etc.

. A three-tier system design consisting of Display Stations, Data

Acquisition Computers and I/O subsystem has been adopted to

improve the reliability and availability. This makes the different

subsystems to be hardware independent on each other. A high speed

Ethernet LAN (Local Area Network) is used for communication

between the subsystems.

10 of the Display Stations are Utility CRTs (UCRTs) and the

remaining 2 are Alarm CRTs (ACRTs). They are Intel 80386

microprocessor based systems doing most of the user interface job.

These systems are having high resolution (Super VGA-1024 x 768

pixels) 19” monitors, which give a good pictorial representation of the

data.

Data Acquisition Computers are based on Intel Pentium which UNIX

SVR 4.2 as the Operating system. Both the DACs work in dual

redundant hot standby mode. They mainly acquire the data from I/O

subsystems and other Computer Systems like PLC, DPHS, RADAS

etc. and pass the required data to display stations. They do the

logging of the history data and take care of the printer tasks. They

also do the network management of both the LANs. They also direct

the I/O systems to govern the field outputs as required.

I/O subsystems are Motorola 68020 CPU based systems. Each I/O

subsystem has 2 CPUs working in dual redundant mode. They mainly

do the scanning and alarm checking of the field inputs connected to


33

them and pass on the data to DACs. They also change the field

outputs as per the directive from DACs.

The network topology is designed in such a way that a single break

anywhere in the network (broken cable or failed n/w equipment) will

not result in a collapse of the total system, but will allow the system to

continue to work at a degraded level. The significant aspects of the

designed network topology are:

a). Thicknet cable is used as it is much more rugged than the

thinnet cable.

b). Transceivers that are used to connect different nodes to the

network are piercing type tap boxes, which facilitate the

connection without a cut in the cable.

c). A significant component in the topology is “Repeater”.

Repeater is an active component that can be used to connect

different cables of networks. It isolates the remaining network

from a fault in any of the other cables. This has given rise to a

fault tolerant network. The network is divided into 4 parts each

connecting ¼ of the system. The various failures considered

their effect is described below:

i). If any transceiver or the cable connecting the transceiver

to the node fails, only that node will fail & reset of the

system will continue to work as usual. If the connection

with the Master DAC fails, the hot standby DAC will

take over and the system will not be affected.

ii). If any one of the cables of LAN1 - a, b, c or d fails than

¼ of inputs/outputs (of the nodes connected to that part

of the network) will be lost & the system will continue to


34

work with 75% of inputs/outputs data. If more than one

cable fails, COIS still will work with the reduced

capacity accordingly.

iii). If any one of the cables of LAN-2 – a, b, c or d fails, then

¼ of display stations will not be available. Display

stations are connected in these four cables in such a way

that CRTs on adjacent Main Control Room panels are

connected to different cables and will not fail

simultaneously. ACRTs are connected to different

repeaters and hence will not fail simultaneously. If more

than one cables fail, COIS will still work with the

corresponding reduction in display stations.

1.1 Inputs/Outputs

There are various types of plant inputs to the COIS viz. analog inputs

and digital inputs. Each plant input is also referred to as “point” the

COIS also provides voltage free relay contacts as outputs.

Analog inputsThere are 1256 analog inputs to the COIS. These include about 10%

spare points. The approximate distribution into different categories is

as follows: -

RTD’S 392 points

Thermocouples 16 points

Volts/Current Inputs 824 points

Thermistor 8 points


35

Exact details of description, input range, alarm priority, process range

and processing required etc. for each analog input are available in the

COIS analog input..

For the current inputs the terminating resistors (of value as specified

in analog input list) are a part of the COIS. Among the 824 voltage or

current inputs, any number may be voltage input. Thus all of these

824 inputs can be arranged to take a current or voltage input. Input

impedance of voltage inputs is greater than 1 Ohm. Linearisation and

lead resistance compensation wherever necessary, e.g. for RTD and

thermocouple inputs, will be performed by the COIS. All RTD inputs

will use 3 wire RTDs in the field. Provision will be use 3 wire RTDs

in the field. Provision will be made to terminate a 3 wire RTD at each

RTD input point.

Cold junction compensation for thermocouple will also be provided

by the COIS. These analog inputs are numbered in the range of 000 to

1299.

1.1.1 Digital (contact) Inputs

There are 1136 digital inputs of contact type. There are two types of

contact inputs as follows:

1.1.1.1 Voltage free contact inputs: There are 736 voltage free

contact inputs. These contacts represent alarm or status inputs.


36

1.1.1.2 Shared contact inputs: There are 400 field contact inputs

which are shared between the window Annunication system (WAS)

and the COIS. These contacts represent alarm events.

1.1.2 Digital (Voltage level) Inputs

There are 656 voltage level inputs representing the status of valves

(open or closed).

State Voltage

Level 0 Between 0 volts and 2 volts

Level 1 Between 40 volts and 48 volts.

The input impedance presented by COIS to these inputs will not be

less than 10K ohm.

1.1.3 Input Data from other computer systems

COIS receives data from other computers viz. Digital Recording

System (DRS), Radiation Data Acquisition System (RADAS),

Electrical DAS and PLC’s via LAN thorough gateways. In the COIS,

these input parameters are numbered as follows:

(1) Radiation Data Acquisition System:

Analog points: 3601 to 3799

Digital points: 3001 to 3999

(2) Electrical DAS:

Analog & contact points: 7001 to 9499

(3) PLC’s Digital points representating : 1301 to 1999

Status of hand switch position & 5901 to 5999

(4)Digital Recording System (DRS)

a) Normal/Disturbance Analog inputs : 2801 to 2899

b) Visicorder Function Analog inputs : 2901 to 2999

c) Contact inputs of ESR function : 9501 to 3499


37

d) Dual Process Hot Standby : 9501 to 9699

Analog inputs

(5)Other Computer Systems

a) Analog inputs : 9701 to 9899

b) Contact/digital inputs : 9901 to 9999

Note: There are no physical inputs corresponding to these points.

Values of these points are provided to the COIS periodically by the

above systems. For all displays and logging functions except alarm

functions, these points are treated as the field COIS inputs.

1.1.4 The COIS will also provide 224 outputs of voltage free relay

contacts. Ten or these contacts are used for Fuel Failure Monitoring

function described in Sec. testing function described in section 5.11.

The remaining contacts will be used for miscellaneous purposes like

giving time synchronizing pulses to other computer based systems,

annunciation of the failure of the COIS, indication of which computer

system is faulty etc.

1.1.5 Ethernet

The COIS will provide Ethernet LAN connectivity for connection to

other computer systems.

1.1.6 The COIS will also provide, on operator’s demand, processed

data outputs called calculated analog variables, for e.g. selected

channel differential temperatures and DNM detector outputs etc.

These variables will be numbered in the range of 6001 to 6699.


38

1.2 Accuracy, Noise Rejection, Contact Debounce and Isolation.

1.2.1 For all digital and analog inputs, a very high isolation between

the transducer circuit and the COIS is provided to avoid problems in

the transducer circuit due to ground faults etc.

1.2.2 Overall accuracy of analog data acquisition for any point will

be 0.25% of span or better. This accuracy will be maintained even in

the presence of common mode noise (Max.) + 15V d.c./50Hz. on the

input. A low pass filter will be provided on each analog input to

suppress normal mode (predominantly) 50Hz. noise. Protection will

be provided against following conditions for different categories of

analog input.240V AC (Max.), 50Hz common mode voltage on any

thermocouple inputs.

For RTD inputs: Depending on RTD bridge excitation

network

Or + 50V d.c./a.c. (Whichever is more) common mode or

normal

mode voltage.+ 250V d.c./a.c. common mode or + 50V

dc/ac

normal mode voltage on any other type of analog inputs.

1.2.3 Processing of potential free contact type inputs will not be

affected even under a max. common mode voltage of + 15V d.c./50Hz

a.c. on any input. Beyond 15V, protection is provided for a max. +

250V d.c. /a.c. common mode voltage. For providing the above

common mode voltage capabilities, opto-isolators are used for digital

inputs.


39

1.2.4 In case of digital (contact) inputs and digital (voltage level)

inputs input status changes lasting less than 50 milliseconds will be

ignored by the system.

Note: Sampling interval for all analog & contact inputs will be

adjustable to any of the following: 5 sec. 10 sec., 30 sec., with the

normal interval specified in the above table. This adjustable could be

on an individual basis or on group basis (Group = type of input).

For all analog inputs, five samples will be taken within the sampling

interval and the average of these five samples will be taken as the

value for that sampling interval.

There will also be a provision for averaging over last five sampling

intervals for selected number of analog inputs (max. 100Nos.). These

average values will be used in all displays and printouts.

1.3 Alarm Function

The computer system will check some of the analog inputs and almost

all of 736 digital alarm inputs in the 656 Digital (voltage) inputs for

alarm events. Many analog points are only for periodic logging and

BG display etc. and are not checked for alarm at all. And some points

have to be checked for alarm at all. And some points have to be

checked for only low alarm limit or only high alarm limit i.e. both

alarms are not required. Some COIS points are inhibited from

reporting to ACRTs as alarms, i.e. these points are not displayed on

the ACRTs when the status of these points changes. But this does not

prevent them from displaying their status in BGs, tabular trends etc.

The remaining point will have both low and high alarm limits. Some

of contact (digital) inputs are for status monitoring only and will not


40

be checked for alarm. The remaining points will be checked for

alarms. The frequency of alarm checking will be same as that or the

point for data acquisition. The alarm events are defined as follows:

1. An analog input going above a high limit (HL) or falling

below a low limit (LL) or a digital input sent to alarm state

since previous scan is referred to as an “Alarm” occurrence.

2. Analog input returning between its high and low point or

a digital input going to normal state since previous scan is

referred to as “return to normal” occurrence. This is also

referred to as ‘Normal’ occurrence is short.

As for as ACRT display or output on alarm printer are concerned,

analog points will be limited to only “low”, “High” and “Bad” states.

BL (Bad Low) than lower end of span Lost. Such additional alarms

will store in the computer memory and arranged as CRT concealed

alarm pages for display purposes.

Capacity for such 100 additional alarms is provided. A suitable

audiovisual indication for the alarm in the concealed pages is

provided. Operator will be able to call up for display any of the ACRT

concealed alarm pages on any of the two ACRTs or on both ACRTs

will be different and independent. Provision will be made for scrolling

up or down (one line at a time) of ACRT display. Latest ‘alarm/’

return to normal’ message line will also be displayed on the last line

of all the ACRT pages. Provision will be made so that the operator

can retain on the screen the most important/of immediate

interest/relevant alarms only on the screen and put the rest of them in

concealed alarm pages. Provision will also be made to list the alarms

on any UCRT for a selected USI or a range of USIs keyed in by the


41

operator. This is called “alarm” display management”. Each input

point will be given a priority number of 1 or 2. Inputs with priority of

2. It will be possible for the operator to call up the summary of

existing alarms on any UCRT. It will also be possible to call this

summary as total or only of alarms with a priority 1 or only of alarms

with a priority 2. Provision will be made to list all alarms for a

selected USI or range of USIs keyed in by the operator.

Operator will be able to ‘tell’ the COIS any analog/digital inputs

which are to be ignored (i.e. as if those do not exist) for alarm

function. Such ‘ignored’ inputs will be resumed automatically in 30

minutes or whenever desired by the operator, whichever is earlier.

Such operator commands will get immediately logged on Alarm

Printer.

The system will maintain a list of such “alarm-disabled points. It will

be possible to add/delete points in this “disabled list”.

The COIS will have the following two codes of alarm processing.

1) Mode 1: Under this mode repetitive alarms are suppressed.

2) Mode 2: Under this mode, repetitive alarms are also reported

(without any suppression) lost. Such additional alarms will be

stored in the computer memory and arranged as CRT concealed

alarm pages for display purposes.

Capacity for such 100 additional alarms (i.e. approx. 5 concealed

alarm pages) is provided. A suitable audiovisual indication for the

alarm in the concealed pages is provided. Operator will be able to call

up for display any of the ACRTs or on both ACRTs i.e. the pages

selection keys for both the ACRTs will be different and independent.

Provision will be made for scrolling up or down (one line at a time) of


42

ACRT display. Latest ‘alarm/’ return to normal’ message line will

also be displayed on the last line of all the ACRT pages. Provision

will be made so that the operator can retain on the screen the most

important/of immediate interest/relevant alarms only on the screen

and put the rest of them in concealed alarm pages. Provision will also

be made to list the alarms on any UCRT for a selected USI or a range

of USIs keyed in by the operator. This is called “alarm display

management”. Each input point will be given a priority number of 1

or 2. Inputs with priority 1 being more important than those with a

priority of 2. It will be existing alarms on any UCRT. It will also be

possible to call this summary as total or only of alarms with a priority

1 or only of alarms with a priority 2. Provision will be made to list all

alarms for a selected USI or range of USIs keyed in by the operator.

Operator will be able to ‘tell’ the COIS any analog/digital inputs

which are to be ignored (i.e. as if those do not exist) for alarm

function. Such ‘ignored’ inputs will be resumed automatically in 30

minutes or whenever desired by the operator, whichever is earlier.

Such operator commands will get immediately logged on Alarm

Printer.

The system will maintain a list of such “alarm-disabled points. It will

be possible to add/delete points in the “disabled list”.

i) Mode 1: Under this mode repetitive alarms are

suppressed.

ii) Mode 2: Under this mode, repetitive alarms are also

reported (without any suppression). The operator through a

password can select alarm-processing mode 1 or 2.

1) Alarm processing under mode 1:


43

Under some abnormal field conditions, some of the points

(analog/contact) may oscillate between alarm and normal states

and hence any cause large number of alarm/normal messages

on the alarm printer & ACRT’s. Hence, it is required that not

more than six message (status changes) are generated by any

point in any quarter of an hour. For this purpose, the COIS will

set the “status change” count of each to zero, every quarter of

an hour. Any point which changes status (from normal to

alarm/bad or alarm/bad to normal etc.) 6 times in any quarter of

the hour, will be automatically disabled from alarm scanning

for the remaining part of the quarter hour. But the COIS will

freeze the status only with alarm/bad status i.e. if 6th message is

normal message, the COIS will disable alarm scanning of the

point after 7th message (i.e. Alarm message) is reported. This

period for checking max. NO. Of alarm generation will be

programmable between ¼ of an hour to a selected period will

also be adjustable between 4 and 10.

Alarm processing under mode 2 :

All alarm messages are reported without any suppression.

1.3.1 There are about 400 Nos. digital inputs (shared input contacts)

which are scanned only for the logging of their status changes on the

alarm printer and mag. Tape cartridge/disk cartridge (i.e. CRT display

and audio ann. Is not required for these). (Note: These are the window


44

annunciator points numbered in the range of 4001 to 4999). These are

scanned every one second.

1.3.2 Latest Alarm Message Display Function on UCRTs

The latest “Alarm”/”Normal” message displayed on the ACRTs will

also be displayed on the bottom-most line of all the UCRTs also. No

flashing of the alarm message or any audio is required on the UCRTs.

However, successive alarm message will be displayed in alternate red

and pink colours on UCRTs (e.g. first alarm message in red colour,

second in pink colour and third in red again and so on). Normal

message will always be shown in green colour. Any alarm/normal

message will be continuously displayed until another alarm/normal

message is generated to replace the previous one. Operator will

provide a facility to switch off this alarm/normal message display on

any UCRT whenever required.

1.3.3 Valve Status Monitoring

There are 656 Nos. voltage level inputs representing the valve status

(open or closed). These inputs are scanned once every 1-second for

displaying the actual status of the valves in the Mimics and for

logging their status changes on printer used for alarm logging. Status

changes are recorded on magnetic tape also.

Each valve to be monitored for its status on COIS will have one or

two inputs connected to COIS. These are voltage level inputs with two

levels, i.e. at 0 volt for 48 volts dc. Voltage level inputs are normally

taken across the indicating lamps corresponding to the valve. When

the indicating lamp is ‘ON’ indicating valve “Fully open” or “fully

closed”, the input to COIS is +48V DC, otherwise it is zero. If there


45

are tow inputs corresponding to a valve, the COIS will sense both the

inputs and derive the status as follows:

“Fully open” input “Fully Closed” Input Status

48 V 0 V Fully open

0 V 48 V Fully closed

0 V 0 V In intermediate

position

48 V 48 V INST. FAILURE

The COIS will show the actual status in the Mimics appropriately. It

will also log the change of status on the printer accordingly.

For valves having single input to COIS, the status will be either

“Fully open” or Not Fully open” (“Not Fully Closed”).

If there is a change in status the new status will be logged on to the

printer. It may be noted that the “Intermediate” status and “Instrument

failure” status are taken as new status only if it has remained so for

two successive scans. The “Instrument failure” status is treated as an

alarm and would be annunciated on the ACRT.

1.4 Interface to Various other Computer Based Systems

The COIS will also provide Ethernet LAN interface for connecting

the following computer systems. The COIS will receive data from

them as required as per the approved protocol. The data will be


46

available for all the functions described in this design manual except

for alarm function:

a) RADAS

b) Electrical SCADA (EDAS).

c) PLC

d) DRS

e) DPHS

f) CTM

g) PDCS etc.

1.5 General Features:

1. The COIS will be user friendly and the operator will be able to get

the desired information in desired formats on the various UCRT’s in

an interactive manner. Menu driven CRT based dialogue with the

system will be designed. Various menus/indexes/lists of UCRT’s

functions, BG’s, History groups and graphic trend groups etc. will be

displayed on the UCRT on operator’s demand. ‘Help’ facility will be

available at all phases of dialogue. Data retrieval procedures will be

quick and easy. No UCRT will have a blank screen at any time. If no

display is demanded, it will be showing the main menu, so that

operator can quickly select the display of his choice.

2. Process data base will contain flags for each Analog input to

indicate whether any Low Alarm Limit/High Alarm Limit is


47

applicable or not. Software for Alarm processing and various displays

etc. will not call for entry of artificial low/high alarm limit e.g. lower

than lower end of span/higher than higher end of span etc. as this

causes confusion and inconvenience to the operator while studying the

printouts/displays.

4. Microprocessor based and standalone type I/O subsystems are

used. Analog inputs cards are designed for automatic calibration at

regular intervals under software control using precision reference

sources for 25% and 75% scale. Hence, software offset correction is

provided for any drift due too temperature or time.

5. A facility for enabling/disabling “low” alarms in bulk for certain

points with a single command will be provided. These low alarms

mostly correspond to the failure of the sensors.

1.6 Resolution of Values in Numeric/Plot/Bar Graph Form

Resolution of Values (i.e. data such as current value of a process variable, alarm

point, etc.) will in various cases be as follows or better:

Sr.

No.

Type of output Resolution

1 Numeric form 0.1%

2 Graphic Display/Mimic/Bar

Graph/Plot

0.2% or

better

Numerical resolution will be limited to 0.1 to 0.2% of full span. The

operator will not require a better resolution than this and it will

waste the useful area of the screen. Hence numerical displays like


48

21.3245 for a process range of 0-100C will be avoided and an

approximated value 21.3 will be displayed.

1.7 Response Time

Expected response times are described as follows:

1. Initial Display Lag : It is defined as the Maximum Time taken

for the first complete display (static + dynamic parts) to appear on

the display after the operator’s common will not be more than 2

seconds.

2. Time Stamp Lag : It is defined as the maximum time

difference between the real time of a field event occurrence and the

time stamp (Time stamp will be done as soon as the scanning is

done) will in principle be same as the sampling interval.

3. Display Lag : It is defined as the maximum time difference

between time stamping of an event and it being displayed will not be

more than 1 second.

4. Sampling interval : It is defined as the maximum time between

the two consecutive scanning of the inputs

5. Print Lag : It is defined as maximum time difference between

the commencement of the demanded printout and the operator’s

commands and will not be more than 10 seconds.

1.8 Data Storage/Retrieval and Off-line Computer System

1.8.1 Data Storage

The following on-line data will be recorded on the magnetic disk for

last 32 hours:

a) History data

b) Static data base


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c) Changes done in static data base

d) Five sets of DNM data & ECCS Test Data

e) Alarm logging

f) Snapshot of current values of all the points at every shift and as

demanded by the operator.

g) Data and time of recording the data.

This data for the last 24 hours, on default, will be dumped on to the

magnetic tape one in a day at a fixed time, which will be adjustable

on system starting time.

Typically, one tape will be used for a day and previous one month’s

data will be stored (i.e. 31 tapes will be available). Before dumping

the data, system will check, if the tape on the drive is of that day’s

tape. If not, it will ask the operator to insert new tape for dumping

the data.

1.8.2 Data Retrieval

Provision will be made to retrieve the data from the disk (current 24

hours) or from any previously recorded magnetic tape and store it on

a PC (MS-DOS) compatible floppy. Facility will be provided to

select any type of data and in any range (time, usi, etc.).

1.8.3 Off-line computer system

Off-line computer system will comprise of a standard PC-AT and a

printer. It will be possible to read the data recorded on floppies by

the on-line system .The software would also include standard

package like DBASE IV.

1.9 Input Power Supply to the Equipment and Effect of Power

Failure.


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Two independent sources of single phase, A.C. power supply of the

following specifications will be available in the station.

Voltage

RMS Value : 240 Volts + 10%

Steady state variation : + 10%

Transient variation : 0% (for 200m secs)

Frequency

Frequency : 50 Hz

Steady state variations : + 1%

Transient variation : + 5%

During transient variations (upto 200m sec) the COIS will continue

to operate without malfunctioning.

An input a.c. power interruption (total outage) lasting 30

milliseconds or less will not affect the working of the COIS in

anyway.

For all 240V AC loads, both sources of main supply will be

connected through contractors such that failure of any main power

supply will not affect operation of any subsystem. Wherever

duplicated D.C. power supplies are used, separate a.c. main source

will be given to the two D.C. power supplies will be connected to

the load through isolation diodes.

1.9.1 Seismic specification

1. The I/O subsystem equipment will operate satisfactorily during

and after the vibration tests at the following peak accelerations when

subjected to a sinusoidal acceleration for 30 seconds at each

frequency in the given range.


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Peak acceleration I the horizontal axes and vertical axis: 3.5g from 1

Hz to 33 Hz.

1.10 Master Clock Time

Real time clock of the COIS Unit-1 will be used as the master clock

for synchronizing time of various computer systems of the plant.

The COIS will provide a potential free change over contact to each

of the computer based systems for time synchronization with 0.5

second status change at 10.00 hours every day (Normal status will

be resumed at 10:00:00 hours). The contact status change will be

sensed by each of the computer based systems and the time will be

set to 10:00:00 hrs.

1.11 Reliability and Availability

The meantime between failures (MTBF) of the system excluding the

printers, plotters and CRT’s will be 4000 hours or more with an

availability of 99.9% or better. MTBF and availability figures for the

printers/plotters and CRT’s will be 2000 hours and 99%

respectively. The meantime to detect a fault (MTTD) plus the mean

time to detect a fault (MTTD) plus the mean time to repair (MTTR)

will not exceed 1 hour. In order to achieve the high availability

figures, a master and hot standby computer redundancy is employed.

In case of a computer going faulty, its load will be automatically

switched over to the other computer. Displays/printouts active

before the failure of computer, become active automatically without


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operator’s intervention after switching over pertaining to History

will not be lost due to such switchover. To keep the MTTD+MTTR

under one hour, ‘hot repairs’ concept is used.


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Documents

Training Report Npcil Rapp Rawatbhata