Basics of Nuclar Plant Operation

7/29/2019 Basics of Nuclar Plant Operation

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BASICS OF NUCLEAR POWER PLANT OPERATION

In a nuclear power station, heat is produced by the fission of the nuclei, such as those of

uranium, in a reactor. Thus the source of heat energy is the reactor, which is equivalent to

the furnace of a fossil fired boiler. The heat generated in the reactor is transported to the

steam generator by the primary heat transport fluid( PHT) . In the steam generator the hot

primary heat transport fluid heats up the DM water to produce steam for running turbine.

The difference between a nuclear plant and a thermal plant lies in the source of energy.

In thermal power station the source of energy is limited to the best calorific value of the

fuel used. There is a limit of the same. But in a nuclear reactor the source energy is

infinite. There is no upper limit that can be produced in a reactor. The maximum limit of

power is determined by the rate at which the heat can be trans ported from the reactor to

keep it cool.

There are following types of reactor.

1. BWR (Boiling Water Reactor):

DM water is used in this type of reactor. Radioactive fuel remained immersed in DM

water. DM water is directly heated in the reactor by fission reaction to produce steam.

The steam is of radioactive. This steam is used for driving turbine. So utmost safety

measures are required for this type of Nuclear power plant.

2. PHWR (Pressurized Heavy Water Reactor):Heating of DM water to produce steam is indirect. Radioactive fuel remains

immersed in the heavy water called (D2O) called primary heat transport fluid. Heat

transport fluid is heated up in the reactor. Hot fluid is then circulated through the

steam generator tube side. The tube side of steam generator is called primary side. In

the steam generator the primary tubes remain immersed in the DM water. Due to the

heat transfer from the hot fluid, the DM water gets converted in to steam (which is

called secondary cycle). Here the steam is not radioactive.

Prepared by S. Mandal / DGM /TSX/Tarapur site

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PWR (Pressurized Water Reactor):

The working principle of this reactor is same as PHWR with only difference that here

heat transport fluid is (normally heavy water which is very costly) replaced with DM

water. So both in primary and secondary cycle DM water is used .The difficulties to

handle heavy water is reduced to a great extent in such type of reactor.

3. FBR (Fast Breeder Reactor):

Here also the basic principle of operation is same as that of PHWR. The fuel used in

this case is Thorium. Outside the fuel tube there is another tube. Plutonium is packed

inside it. During fission the fast neutron is absorbed by plutonium and get converted

into Uranium. This Uranium can be used as primary fuel for other reactor. The

necessity of this arisen out of less and less availability of natural uranium.

In NPC, Tarapur the reactor used is of PHWR type. In PHWR heavy water is used in

reactor both in shell side and tube side. Heavy water filled in the shell side is called

Moderator and the same in the tube side is called Heat Transport Fluid .There are four

nos of steam generators with one such reactor. Each steam generator is capable of

generating 750 tph (appox) of steam at appox. 250C and 42-kg/cm2 pressures. This

reactor (called Calandria) is like a condenser with 392 nos of hollow tubes. The tubes

are supported on the end shields. They are called calandia tube. Concentric with these

tubes there is another set of tubes. They are called pressure tube/fuel channel. The

annular gap between these two tubes is filled with CO2. Inside the pressure tubes the

radioactive fuel Uranium (U-235) bundles ( 13 X 292 nos ) are placed in concentric

manner. Through the annular space the heavy water(called heat transport fluid) is

circulated at high pressure (at 100 kg/cm2) and high flow by four nos of pumps

(called PCP- primary coolant pump). Because of fission reaction heat transport fluid

in the calandria tube gets heated up. Hot heat transport fluid flows through the tubes

of the steam generators. In the steam generator these tubes are kept immersed in DM

water. DM water exchanges heat with heat transport fluid and gets converted into

saturated steam. This steam is used for driving the turbine.


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CALANDRIA TUBE CO2

FUEL SHEATH FUEL ELEMENT

FUEL CHANNEL HEAT TRANSPORT FLUID

CROSS SECTION OF TUBE


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HEAT TRANSPORT FLUID

FUEL

SHEATH REACTOR TUBE

TUBE

HEAT TRANSPORTFLUID FLOW

REACTOR OR CALANDRIA

REACTOR

VESSEL

FUEL

CHANNEL

M

O

DE

R

AT

O

R

FU

E

L


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HEAVY WATER AS MODERATOR.

Moderator is filled in the shell side of the reactor. Neutron released during the chain

reaction have an average energy of 2 Mev whereas neutron most likely to cause

fission (called thermal neutrons) with energy comparable to 0.025Mev.Hence it is

necessary to reduce the energy of the fission neutrons for transforming them to

thermal neutrons. Moderator actually slowed down the energy of the fission neutrons

by absorbing its energy so that they get transformed to thermal neutrons. Fission of

all variety of neutrons will not occur with the thermal neutrons. Only U-233,U-235 &

Pu-239 fission occurs with thermal neutrons and these are generally used as nuclear

fuel in the nuclear power station.

The moderator is heated up by absorbing the energy of fast neutrons. The total heat

produced in or transferred to the moderator is about 7% of the total heat produced in

the reactor.

When fission occurs, the nucleus divides into two nuclei called fission product. They

are usually unstable and decay into other nuclei. If not controlled, these products can

be hazardous to station personnel. They also absorb neutrons to a grater / lesser

degree and acts as poison to reactor.

Neutron

Neutron gama ray

Neutron

These poison if not continuously removed the load builds up above the normalequilibrium level resulting in so called Xenon transient. It is uneconomical to have

excess radioactivity to overcome this transient. Under shut down/ trip condition from

full power operation, if the reactor is not started within 10-15 minutes, it will be

incapable of starting up to several hours due to accumulation of poison. This delay is

known as poison out time. The poison override time, during which the reactor can be


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U-235

U-

236

Xe-

144

Sr-90


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started after a trip, can be extended by booster fuel rod. But this requires a higher fuel

inventory.

So during operation, moderator gets heated up and contaminated. In order that

moderator continues to function properly, the desired neutron properties must be

maintained. This involves the treatment of moderator such that

a. The temperature is prevented from rise above set value (


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Heat is produced in the reactor by fission of the fissile material and decay of fission

products. Over 6% of total heat is produced by fission decay after reactor trip out / shut

down. Within first 10-20 seconds after trip out, the reactor power id reduced to 6-7% of

full power. Thereafter the decrease of remaining power is slow because of the heat is

produced by fission decay. This heat if not removed will heat up the secondary side DM

water by convection to produce steam. This may cause temperature rise in the primary sid

e to such an extent to rupture / melt the fuel sheath and channels. Under normal condition

when the PCPs are available, there is no problem of circulating the heat transport fluid

through the reactor during fission decay. To meet the emergency condition when PCPs

are not available, standby cooling pumps( shut down cooling pump) are incorporated in

to primary the system .In a similar manner when BFPs are not available the steam

generator is cooled down by emergency BFP during fission decay period.

As in the case of the moderator, heat trans port fluid becomes contaminated with the

corrosion product. The heat transport system is predominantly of carbon steel

construction and a PH of 9.5 11 is required for the system. A purification system is

incorporated to continuously purify the contaminated the heat transport fluid. Since the

purification system will not work properly with high temperature fluid, heat exchanger is

used to cool down fluid acceptable to purification system. The purified fluid then gets

heated up through the Rene generative heat exchanger before being injected into the

system


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MAJOR PRE- COMMISSIOING ACTIVITIES OF NUCLEAR REACTOR AND

STEAM GENERATOR.

Like conventional boiler following pre-commissioning activities are carried out before

putting the nuclear

1. AIR LEAK TEST OF CONTAINMANT

2. HOT CONDITION

3. DRY OUT

4. FUEL LOADING

5. CRITICALITY

6. ASDV &POSRV OPERATION

1. AIR LEAK TEST

The reactor and its auxiliaries in the nuclear plant are housed in the concrete containment.

There are two concentric containments. During normal operation both the containments

remain under sub atmospheric pressure. A fan sucks the air from inside the containments

and discharge it to the atmosphere through the chimney (like ID fan in conventional

boiler). The negative pressure in the inner containment is maintained at about -80mm of

H2O in the shut down accessible area and about -50 mm in the accessible zone. The same

in the outer one is 20 to -30 mm of H2O.Fresh filtered air is inducted into the

containment to maintain the draft. In the discharge duct of the fan there is measurement

of radioactive element of the gases. When the measurement exceeds the limit value, the

plant is shut down.

For ensuring leak proof containment , the inner containment is leak tested with air at

pressure of appox 1.5 kg/cm2 for 4 hrs ( like furnace leak test). After successful leak test

of the same it is cleared for normal operation.


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2. HOT CONDITION

Hot conditioning is the process of forming magnetite layer inside the pipes and internals

of the primary circuit through which the heat transport fluid is circulated..

For achieving the same both the primary side and secondary side of the steam generator

is filled with DM water. High power ( 6 MW each) PCPs ( 4 nos) are kept running so that

the electrical energy is converted into heat energy. The pressure and temperature of the

secondary is raised to rated value (48 kg/cm2 @ 290-300C). The condition is maintained

by circulating the DM water through secondary circuit. Auxiliary BFP is continuously

run for circulating the DM water through the steam generator secondary side. In this

process the DM water of the secondary side gets heated up and converted into steam.

Intermittent/ continuous venting of the steam from the secondary is done to keep the

primary side parameter under desirable condition. To achieve the desired thickness of the

magnetite layer, the hot conditioning activity is carried out from 5-7 days. A sample piece

is kept in the primary side. At the end of process the sample is checked for the desired

thickness of the magnetite layer. This similar to passivation operation in conventional

boiler

3. DRY OUT

After hot conditioning the whole charge of DM water is drained from both the primary

and secondary side of steam generator. Hot dry air is circulated through the primary side

of the steam generator for completely drying out the primary circuit. For this auxiliary

steam at 6-7 kg/cm2 @ 150-180 C is circulated through the secondary side of the steam

generator. Instrument air is circulated / vented through the primary side . The instrument

air gets heated up in the steam generator and dries the prevailing moisture form the

primary side (which remains at post hot conditioning activity).


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4. FUEL LOADING

After ensuring the dryness of the primary side and readiness of the control system for fuel

(fissile material) loading, fuel (normally U-233) is loaded in the pressure tubes of the

calandria. Fuel comes in the form of rod. Total 13 x 292 nos of rods are loaded in the

pressure tubes. Fuel rod as it is much not radioactive and hence can be loaded in the

pressure tubes. The fuel when comes in contact with heat transport fluid i.e. heavy water

becomes radioactive

5. CRITICALITY OF REACTOR.

Criticality of a nuclear reactor is defined as a ratio (neutron multiplication ratio-K) .When

it becomes equal to 1 i.e. K=1 we call the reactor critical. Uranium when comes in

contact with heavy water, starts chain reaction. Unless it is controlled, it will be

hazardous. So when heavy water is charged into the reactor (in the primary side)

Gadolinium in liquid form is mixed with it. Gadolinium is called poison because they

absorb most of the free neutron and hence prevent chain reaction.

To achieve criticality, these poison are slowly removed through the moderator

purification system. Usually this activity takes about 60-72 hrs. While removing the

poison from the moderator, free neutrons will be available for chain reaction. First

criticality of reactor is the point when sustained chain reaction is achieved ( with energy

in the level of micro watt) .For useful energy to be obtained from reactor , its criticality is

to be maintained at higher level where the heat transport fluid is heated up so that when

it is circulated through the steam generator , it will boil the DM water ( in the secondary

side) to produce steam.

To raise the reactor power the neutron absorption by poison is to be minimized. Also the

neutron loss (through neutron escape) is to be minimized. The reactor power is

maintained at desired level by either absorption of neutron (to reduce the value of K < 1)

or by neutron multiplication (by increasing the value of K > 1).


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MAIN STEAM OUT

STEAM DRIERS

MOISTURE

SEPERATOR

STEAM GENERATOR

STEAM GENERATOR


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FEED WATER

HEAT TRANSPORT FLUID HEAT TRANSPORT FLUID


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ASDV AND POSRV OPERATION

First operation in a nuclear plant after achieving criticality is testing the availability

ASDV (atmospheric steam dump valve) operation. This operations like safety valve

floating in conventional boiler. Steam dumping continues with low reactor power (around5%). Steam generator outlet pressure is raised by CSDV operation (in manual mode) so

that the boiler pressure signal reaches to the set point (above 17 mA). Auto operation of

each ASDV is tested one by one. Manual operation of POSRV (pilot operated safety

relief valve) before clearing the reactor for normal operation.

BASIC OPERATING PRINCIPLE OF NUCLEAR POWER PLANT ( ref scheme 2)

Saturated main steam (dryness fraction 0.95) is produced in the steam generator. This

steam flows through the main steam pipe (left /right) to HP turbine for expansion. From

the HP turbine it flows to the MSR (moisture separator and reheater) via CRH lines

(left /right). In the MSR the bleed steam from main steam and extraction steam from HP

turbine reheats CRH steam. The reheated steam flows to the LP turbine 1&2 via the HRH

lines for further expansion in the LP turbines.

During start up for achieving the main steam parameter, a part of the main steam is

branched off from main steam (left and right side) and dumped to the condenser through

the CDS (condensate steam dump) valves (4nos two for each LP turbine). In case of

sudden load reduction part of main steam flows to condenser through CSD valves and a

part of HRH steam flows directly to the condenser via the four nos (two for each LP

turbine) LPBP valves bypassing the LP turbines. To keep the reactor power in 100% , if

required four nos of ASDV( atmospheric steam dump valves) will also open if the main

steam pressure increases beyond set value even after the opening of CSD and LPBP

valves. There are 12 nos of POSRV (pilot operated safety relief valves) three in each

steam generator outlet. These valves are like ERV in our conventional boiler. They are

opened from control room manually when required.


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There is LP & HP heater like conventional thermal power plant. Working principle for

them is identical

The basic philosophy of nuclear power plant is to keep the reactor as cool as possible

(during running /shutdown). As stated earlier source of energy in nuclear power plant is

infinite. During operation heat produced in reactor must be evacuated. There are controls

to change the reactor power i.e. heat produced in reactor. It has been observed that even

with most stringent control, rapid changing of reactor power to zero level is not possible

within short period( because of decay heat this is something like our CFBC boiler

where there remains a heavy heat load even after tripping of boiler) . So to keep the

reactor working, either the plant has to operate at its maximum power continuously or the

excess steam has to be diverted temporarily to other circuits (when the turbine load

changes) to allow the time to the control system to bring the reactor power down to new

operating point. Hence it comes to the point that reactor power i.e. the steam produced in

the steam generator must not be disturbed frequently. To achieve this condition almost all

nuclear power plant operates with SGPC (steam generator pressure control) principle i.e

turbine follow mode.

20mA SIGNAL FOR ASDV

17

PROPORTIONAL SIGNAL FOR CSDV

12+ CONTROL DEVIATION FOR TURBINE

8

-- CONTROL DEVIATION FOR TURBINE

40

SGPC SIGNAL


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20 mA

17mA

CSDV STOP AND CONTROL VALVE

12mA 6

4 WATER INJECTION VALVE

SGPC SIGNAL

0mA

SIGNAL TO MAIN STEAM MAXIMUM PRESSURE CONTROL

20mA

TURBINE CONTROL VALVE OPEN

12mA

12mA STOP

TURBINE CONTROL VALVE CLOSES

4mA 4mA

SGPC SIGNAL

0 mA

SIGNAL TO TURBINE CONTROL


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The requirement for highly flexible operational capabilities for large nuclear plants has

made equipment for co-coordinating the various modes of operation an indispensable

pre-requisite for economic unit operation. The governor valve and the steam dump valves

with their control and instrumentation fall in to this category.

In nuclear power plants, the main steam pressure before ESV is an important parameter

controlled by the main steam maximum pressure controller (MSMPC) during

certain operating condition and accident condition.

Since the plant is operated in turbine follow mode mode, so if the turbine cannot take

up all the main steam (during start up, load rejection etc.) the condenser steam dump

valves (CSDV) must discharge the excess steam into the condenser. The CSDVs are

controlled by the main steam maximum pressure controller.

The function of the main steam maximum pressure controller is to support the steam

generator pressure controller (SGPC) to maintain the steam pressure at a pre determined

load related level by actuating though CSDVs. Operation and function of CSDVs are

alike to the LPBP valves of thermal plant. They are operated during start up for

improving steam quality, during load rejection to dump excess steam to condenser. Here

also spray water is admitted into the steam before being dumped into the condenser.

The steam pressure set point is received from the SGPC. Two signals are received

through a MAX selection. This signal is adapted to the required set point for the valve lift

controller of CSDV with the help of a curve generator.

Binary signal for automatic crash cooling are received in three channels. These signals

are selected in 2V3 selection and switches over the set point to maximum, which makes

all the CSD valves to open full.

During load operation of the turbine generator set, the lift set point control loop takes

over the control of the set from the speed set point. The function of the lift set point

controller allows valve lift and thereby turbine load to be increased continuously up to the

point at which the SGPC pressure deviation signal becomes zero. This function consists

of lift set point adjuster and ON/OFF logic. The lift set point controller the limitation set

manually by the maximum valve lift set point adjuster determines the valve lift set point.

This is similar to the conventional load controller of thermal set with initial pressure

control in operation.


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If the actual pressure falls below a certain preset value (+/-10 kg/cm2) from the set point ,

the minimum pressure controller takes over the control of the turbine and the lift set point

controller is deactivated. If the actual pressure improves again, the minimum pressure is

deactivated and control of the turbine is transferred to the lift set point controller.


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Basics of Nuclar Plant Operation