PWR and All Components

8/9/2019 PWR and All Components

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Lecture delivered by Dr. Naseem Irfan(i) of xv

PAKISTAN INSTITUTE OF ENGINEERING AND APPLIED SCIENCES

INTRODUCTION TO NUCLEAR SCIENCE AND TECHNOLOGY

(Lecture Set– 3)

AN INTRODUCTION TO NUCLEAR MATERIALS COMPONENTS AND POWER PLANT COMPLEX

A self-sustaining controlled nuclear chain reaction is a source of nuclear energy. A nuclear reactor is a device in which

nuclear fission is produced under a self-sustaining controlled nuclear chain reaction. It may be looked as a sort offurnace which burns fuel like U-235, U-233 or Pu-239 and products, in turn, many useful products like neutrons, heat,

radio-isotopes etc are formed.

The general components of nuclear reactors are 'Reactor core' containing natural or enriched fuel, 'reactor coolant',

'reactor moderator', material for nuclear reaction control, reactor shielding and the reflectors etc. Core is the main

part of the reactor, which contains the fissionable material called the reactor fuel. The nuclear reactions takes place in

the core and huge quantity of heat is generated. This heat may lead to mechanical destruction of the fuel unless it is

absorbed by some fluid and is taken away. This fluid is called coolant because it cools the fuel elements. Thus reactor

coolant performs dual function. The first is that it transports the large amount of heat from the core to some heat

exchanger where this heat can be utilized to produce steam. Secondly while transportation it sucks heat from the fuel

material and carries it to heat exchanger thus keeps the fuel assemblies at a safe temperature to avoid their melting and

destruction. The function of a reactor moderator is to slow down the fast neutrons ( MeV range) to slow neutrons (ev

range) and this is done in a fraction of second and thus the probability of reaction is increased. The slowing down of

neutrons may be done effectively by light elements such as compounds containing hydrogen, deuterium, carbon or

beryllium. Graphite, heavy water or beryllium can be used as moderator with natural uranium. The ordinary water is

used as moderator only when enriched uranium is used as a fuel.

THE NUCLEAR POWER REACTORS

Pressurized Water Reactors (PWRs)

One of the popular reactor types of the power reactors is the Pressurized Water reactors (PWRs). It comes under the

classification of light water cooled, light water moderated, low enriched uranium, thermal power reactor. The Chashma

Nuclear Power Plant i.e. CHASNUPP is one of such type. PWRs operate on the principle of indirect heating cycle. This

means that steam is not produce directly from the core, instead, heat is transported from the core through a primary

coolant (i.e. light water) residing in the primary coolant loop. Figure 3.1 shows a simplified schematic diagram of a

PWR plant showing primary, secondary and tertiary loops. The primary loop consists mainly of the reactor pressure

vessel, reactor coolant pump and a presurizer along with a steam generator which connects the primary and secondaryloop. The reactor pressure vessel (RPV) contains the nuclear reactor core in which fission reaction occurs.

Simplified Operational Description of PWRs

The primary coolant circuit water flows through the core to remove the fission-produced heat. The junction point of

primary and secondary loops is the steam generator, in which heat from primary coolant is transferred to a secondary

loop and steam is generated. The diagram shows, for simplicity, only one primary and one secondary loop. In actual

practice, in order to increase the power output from the plant while keeping the components down to a reasonable (and

available) size, several loops are attached to the pressure vessel. One, two, three or even four-loop configurations are

utilized as shown in Figure 8.10. The primary system with all its piping and components is enclosed in a specially


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Lecture delivered by Dr. Naseem Irfan(ii) of xv

designed structure called the containment. Thus any radioactivity that may leak from the fuel into the primary water is

isolated from the environment and hence primary circuit acts as a barrier against spread of radioactivity in the

environment.

The name of the PWR is derived from the fact that water in the primary loop is kept at a pressure of about 2250 psi ( or

15.5 Mpa), and a maximum temperature of about 600oF (315

oC). Before discussing how the PWR works, it is beneficial

to understand why reactor coolant pressure is important in PWR. One of the important properties of water is its boiling

point, which at normal atmospheric pressure is 100oC. However, at higher pressures, the boiling point is higher. For

example, at the typical operating pressure of a PWR, namely 2250 psia, the boiling point of the reactor coolant is about

345oC. By maintaining the high pressure, it is possible to prevent reactor coolant from boiling and forming steam within

the reactor vessel may be avoided. In this way a higher heat content is retained in the coolant.

The reactor coolant, i.e. pressurized light water, enters the reactor vessel and flows through the core where it absorbs the

heat due to fission reaction. It leaves the reactor vessel and then flow through the tube side of the vertical U-tube steam

generator where heat is transferred to the secondary system. Reactor coolant pumps return the reactor coolant to thereactor vessel and supplies the dynamic head required for circulation. The presurizer controls the system pressure

transients and keep the pressure and system expansion volume within designed limits and range.

When primary coolant passes through the steam generator, it heats the secondary water-cooling loop and saturated steam

is produced in the shell side. This saturated steam generated in the shell side flows upward through moisture separators

and steam dryers, which reduce its moisture content to less than 0.2%. This steam is used for transporting thermal

energy form the steam generators to the turbine, where it is converted to mechanical and electrical energy. Energy is

converted as the steam expanded through the nozzle and blades of the turbine.

Since it takes substantial pumping power to transfer the steam, it is condensed and changed to water phase. The

condenser is a large heat exchanger connected to the low-pressure turbine exhaust stage. The cooling water passes

through the tube with the condensing steam flowing over their outer sides. The condensate from the condenser after

preheating through the regenerative feed heating cycle is fed back to the steam generators. Sufficient feed-water storage

capacity is maintained within the condensate/feed water systems to accommodate the expansion and contraction arising

from the thermal and pressure effects on steam generator fluid inventory and condensate feed system during the load

changes.


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Lecture delivered by Dr. Naseem Irfan(v) of xv

DESCRIPTION OF THE MAJOR PARTS OF A PWRS

The Reactor Pressure Vessel (RPV):

The reactor pressure vessel is the central component of nuclear steam supply system. It contains the reactor core with

hundreds of fuel assemblies and control rods regulating the fission process. It also contains many other components to

direct the flow of the reactor coolant to and from the core.

As mentioned earlier that the pressurized-water reactor (PWR) operates at

conditions under which the water flowing through the reactor does not boil

because the system pressure is kept high, about 2250 psi, to achieve high

temperature (around 300oC) without boiling in the primary system. The

main function of a pressure vessel is to contain the cooling liquid at very

high pressure and stops the radiation from coming out. In order for a nuclear

power plant to operate properly, all of the internal components contained

within the reactor vessel must be positioned precisely and remain in position

in spite of large forces caused by large temperature differences and high

reactor coolant flow rates.

In general the reactor vessel is made up of carbon steel with its internal surfaces clad with stainless steel [Fig 3.2]. The

control rods drives operated the control assemblies from the top whereas the instrumental thimbles are usually inserted

from the bottom. Depending upon the operating pressure and mechanical loads, the thickness of the reactor vessel

ranges from 10cm (4in) to 35cm (14in). Diameter of RPV ranges from 3.0m (10ft) to 4.9m (16ft) and height ranges

from 10m (33 ft) to 13m ( 42 ft). A typical PWR reactor vessel is shown in figure 3.3 without internal structure for

simplification and steps-wise understanding. The reactor coolant inlet and outlet penetrations are through cylindrical

shell. The control rod penetrations are through the vessel head; the core instrumentation penetrations are usually

through bottom. The typical PWR vessel shown in the figure 2.3 has its top closure about 39ft (~12m) high, about 21ft

(~ 6.4m) in diameter and weighs about 500 tons. The wall thickness of the vessel is about 9 inches (~24cm) below the

reactor coolant inlet and outlet nozzles and is about 14 inches (~ 36cm) above these nozzles. The inlet nozzle has a

diameter of 28 inches (~70 cm) and of outlet nozzle is about 36 inches (~90cm).

Stepping towards a bit more detail, figure 3.4 shows a PWR reactor vessel with its major reactor internals. The different

PWR manufacturers use different internal arrangements; only one of the several designs is shown. The flow path ofreactor coolant through the vessel and internals such that it flows along the internal periphery and reaches the bottom.

From bottom it is pushed into the core through the distributor in order to cool the entire fuel element. Starting from the

top of the vessel, the first major reactor internal is the upper support plate. This plate aligns and supports the control rod

guide tubes. The control rod guide tubes guide the control rod shafts and control rods into the reactor core. Hanging

from a lip just below the upper support plate are the core support cylinder (also called core support shield) and the

plenum cylinder. The core support cylinder actually carries the weight of the core and the lower internals package,

including the core barrel, the upper grid, the lower grid, the thermal shield and the flow mixer (or flow distributor). The

core support cylinder also directs incoming reactor coolant flow downward around the outside of the core and helps

direct the hot reactor coolant leaving the core into the outer nozzles.

Figure 3.2Simplified Reactor Vessel (RPV)


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Figure 3.3 Typical dimensions of a RPV (PWR).


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The Nuclear Reactor Core:

The PWR reactor core, which is enclosed in reactor vessel and supported by reactor

Figure 3.4 Typical Internals of a RPV (PWR).


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Lecture delivered by Dr. Naseem Irfan(viii) of xv

internals, is the heart of a nuclear power plant. It is the source of heat from which the

basic nuclear power cycle starts. It consists of fuel assemblies and all components,

which can be inserted into fuel assemblies to effect reactor power, power distribution

and neutron flux distribution. The PWR core contains the nuclear fuel, which is 2-4%

enriched uranium dioxide UO2. This UO2 is formed into small cylindrical pellets

about one-third of an inch (~ 8-10 mm) in diameter and about three-quarters of aninch ( ~10-18 mm) high. The pallets are sintered, machined and nearly 200 of these

pellets are stacked inside a long thin tube as shown in the figure 7.8 to form a fuel

rod. Specifically the space formed at the top of the pellet stack is called the fuel rod

plenum. It provides a volume for the collection of gases (mostly Krypton and

Xenon) formed in the fissioning process. During assembly the pellets are stacked in

the cladding to the required fuel height. The compression spring is then inserted into

the top end of the fuel tube and the end plugs pressed into the ends of the tube and

welded [Fig. 3.5].

The fuel rods in a PWR are arranged into fuel assemblies, which are sometimes also

called fuel bundles or fuel elements [Fig 3.6]. There are two types of fuel assemblies

in a PWR; those that contain only the fuel rods; and those, which also contain control

rod assemblies. A typical PWR fuel assembly consists of a square matrix of 15 x 15

or 17 x 17 rods as shown in the figure 3.7. The fuel assemblies or bundles along with

reactivity control component and monitoring instrument form the reactor core. About

40,000 to 50, 000 fuel pins/rods are there in a reactor core.

Control rod assemblies for Westinghouse and B&W reactors are usually of the spider

design [Fig 3.8]. Spider control rods use Ag-In-Cd in the form of extruded single

length rods which are sealed in SS tubes. The control element assemblies are guided

within the core by guide tubes which are integral parts of the fuel assemblies. The

spider assembly can be seen to be in the form of a central hub with radial vanes

containing cylindrical fingers from which absorber rods are suspended. The overall

length of the rods is such that, when the assembly is withdrawn through its full travel,

the tips of the absorber rods remain engaged in the guide thimbles so that alignment

between rods and thimbles is always maintained. All of the fuel assemblies in a PWR

are of the same general construction. They are all of same height about 14 ft (~4.25m)

high with a square cross-section of about 8 inch (~21cm) on each side. However, the

enrichment of the nuclear fuel in the fuel rods in one group of assemblies can be

different from the enrichment in another group

Figure 3.6 Stacking fuel pins

Figure 3.7 PWR fuel assembly


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STEAM GENERATOR:

The steam generators are large PWR components in which heat from

the primary circuit is transferred to the secondary with the production of

Fig 3.8 A spider cluster with fuel


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steam. There are two basic steam generator designs: the U-tube and the

once through steam generator.

The operation of the U-tube steam generator is shown in figure 3.9.

This design is by far the most widely used. In it the hot water coming

from the reactor enters at the bottom, fills the one-half space of the

lower chamber, passes upward through thousands of U-shaped inconel

tubes, returns to the other half of the bottom chamber, and returns to the

reactor. The secondary-loop water enters through feedwater inlet into

an annular space between the shell and cylindrical skirt inside the steam

generator, moving first downward in the annulus and then upward in the

space of the tube bundle.

Because of the lower pressure (about 1100 psi) of the secondary loop,

water boils to create steam, which moves upward. The upper part of the

steam generator, called the upper shell, is of a larger diameter than the

lower shell, and contains equipment to remove droplets of moisture

from the steam and to provide drier steam at the outlet, which is located

at the top of the vessel. A large fraction of water, which does not

vaporize, moves back down through the annular space and repeats the

process. Thus, a large amount of water recalculates in the steam

generator.

Based on there respective function, the lower portion of the steam

generator is called the evaporator section and the upper portion is called

the steam drum. The U-tubes are inserted in a thick plate (called the

tube sheet) at the bottom of the component and are held apart and

stiffened along their height, by several thinner circular plates.

Sometimes a steel latticework called an ‘egg-crate’ is used. Additional

metallic holders are used to provide rigidity to the tube bundle and to

minimize vibration induced by the flow of water.

In the popular type of steam generator designs such as the one shown in figure 3.10, feedwater enters at the lower part

of the component. These steam generators include a pre-heater section on the shell side of the feedwater inlet. The

cutout view of the steam generator in this figure shows the U-tubes bundles. The tube bundle supported by the tube plate

and held together by tube supports occupies the lower shell. Moisture separators occupy the upper shell.

Figure 3.9

Schematic diagram showing the soperation of a U-tube

steam generator. Hot water enters from lower left

nozzle, moves upward through the riser section of the U-

tubes and returns through the downcomer section to the

right-hand half of the lower plenum to return to the

reactor. Feedwater moves downward through the

annular space around the bundle, moves upward

through the bundle section, and boils into steam.

Moisture is removed from the steam through the devices

at the top section of the steam generator and saturated

steam of high quality exits from the top of the vessel.

Water droplets fall back and move downward through

the periphery to repeat the cycle.


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Lecture delivered by Dr. Naseem Irfan(xi) of xv

Each loop in a PWR has a steam generator. Each steam generator is a large vessel up to 20m (~63 ft) in height with an

upper shell diameter of about 4.5m (~15ft) and a lower shell diameter of about 3.5(~11ft). The tube bundle incorporates

3260 individual U-tubes, of inconel-600. Inconel is a nickel based alloy with about 76% nickel, about 15.5% Chromium,

8% iron and remaining as carbon.

REACTOR COOLANT PUMPS (RCP):

Each closed loop of the reactor coolant system contains coolant pump

to move the reactor coolant through the loop. It develops the necessary

pressure head to overcome the friction losses that the fluid suffers in

flowing through the core, plena, piping, and steam generator

Figure 3.10 Cutaway and sectional view of a steam generator using U-tubes.


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tubes. These pumps are large vertical, single stage centrifugal pumps.

They are designed to pump large volumes, with high discharge head, at

high system pressures and at high temperatures. Since the coolant

temperature increase in the reactor core is only a modest 28-33oC, a

large volume of water must be circulated to remove the heat generated

in core. Indeed, in a 1200 MWe plant all the pumps handle theenormous amount of about 21.5m

3of coolant per second

(corresponding to about 140 million pound of water per hour).

Each of the primary pumps in a PWR is about 10m (~30ft) high,

requires an electric motor of 10, 000 horsepower, and has a rated

pumping capacity of almost 96000 gallons per minute (6m3/sec). The

inlet is from the bottom and discharge is from the side. In order to

extend the flow of coolant through the core, in the event of a station

blackout, a flywheel is provided on the shaft above the motor. The

stored energy in the flywheel extends the period of coastdown.

Radioactivity of the reactor coolant is the principal problem that makes

reactor coolant pumps different from their conventional counterparts.

Due to this problem, it is important to prevent or strictly limit the

amount of reactor coolant leakage and therefore the current designs

PWRs utilize limited-leakage reactor coolant pumps [Figure 3.11 ].

To start with, it is better to consider first the main parts of the RCP

instead of going into precise details. The reactor coolant pump may be

divided into three main sections. First is the Hydraulic section that is

the lower portion i.e. casing, impeller, diffuser and turning vanes, pump

bearing etc., second is the Shaft Seal Section that contains three seals,

and third is the Motor Section [Figure 3.12]. As the components of the

primary system operating under high temperature and pressure and

having the all-important function to provide circulating water to cool

the core, primary pumps must be designed and manufactured under the

stringent criteria that apply to all primary system components. They are

classified Safety Class I and provisions of section III of the ASME code

apply.

The shafts of these pumps are equipped with properly designed seals to prevent the leakage of the radioactive coolant.

To accomplish this, water form a separate, clean source (CVCS system) is injected into the seal at a pressure somewhat

higher than the primary pressure. The injected water leaks partly inward into the primary system and partly outward

[Figure 3.13].

Figure 3.11 Typical limited leakage RCP

Figure 3.12 Main sections of a typical RCP


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This small outward leakage is easily collected and handled by Chemical and Volume Control System (CVCS). All

plants use more than one primary loop and pump thus number of pumps is redundant. Each of the pumps has more than

enough capacity to provide adequate cooling of the core after reactor shutdown. An air cooled, three-phase ac-induction

motor is mounted vertically on the top of the pump. Figure 3.14 shows details of a PWR reactor coolant pumps, which

would be further explained, in coming semesters in NPPS course.

Figure 3.13 Seal flow diagram in a PWR reactor coolant pump.


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The major components of the primary system PWR discussed in some details in the previous lectures such as reactor

vessel, presurizer, steam generator and primary coolant pump are collectively shown in figure 3.15 to give an idea how

they are collectively installed and work in the nuclear steam supply system (NSSS). It should be noted that there are four

coolant pumps, one for each coolant loop but only one presurizer for the entire system.

Figure 3.14 Internal details of a PWR reactor coolant pump.


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Figure 3.15 Arrangement of the major components of a NSSS (PWR) four loop system.

PWR and All Components

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