Power Reactors Commonly Used in Nuclear Power Plants

Power reactors commonly employed in nuclear power plants are briefly described as follows: 1. Boiling Water Reactor (BWR) 2. Pressurized Water Reactor (PWR ) 3. Gas Cooled Reactor 4. Heavy Water Cooled and Moderated (CANDU TYPE) Reactor 5. Liquid Metal Cooled Reactor  6. Fast Breeder Reactor.

1. Boiling Water Reactor (BWR):

This is the simplest type of water reactor. It has a steel pressure vessel surrounded by a concrete shield. Fuel used is enriched uranium oxide. Ordinary water is used both as moderator and coolant. The steam is generated in the reactor itself. Feed water enters the reactor vessel at the bottom and takes the heat produced due to fission of fuel and gets converted into steam.

This steam leaves the reactor at the top and after passing through turbine and condenser returns to the reactor. Uranium fuel elements are arranged in a particular lattice form inside the pressure vessel containing water. A BWR assembly comprises 90-100 fuel rods and there are up to 750 assemblies in a core holding up to 140 tonnes of uranium.

The secondary control system involves restricting water flow through the core so that more steam in top part reduces moderation. Most of the radioactivity in the water is very short lived (mostly N-16, with a 7 second half life), so the turbine hall can be entered soon after the reactor is shut down.

Advantages:

Advantages of this reactor include a small size pressure vessel, high steam pressure simple construction and elimi­nation of heat exchanger circuit resulting in reduction in cost and gain in thermal efficiency. Overall efficiency is about 33%.

Disadvantages:

1. In view of the direct cycle there is a danger of radioactive contamination of steam, therefore, more elaborate safety measures are to be provided for pip­ing and turbine. These add to cost.

2. Because of the danger of small amounts of fissile materials passing through along with the coolant, more biological protection is required.

3. Wastage of steam results in lowering of thermal ef­ficiency on part load operation.

4. It cannot meet a sudden increase in load.

The 420 MW power stations, at Tarapur (India), consist of two enriched uranium reactors of the boiling water type. These reactors were built with the help of G.E.C. of the United States and became operational on 1.4.1969.

2. Pressurized Water Reactor (PWR):

It is a thermal reactor, using enriched uranium oxide, clad in zircalloy as fuel. A PWR has fuel assemblies of 200-300 rods each, ar­ranged vertically in the core, and a large reactor would have about 150-250 fuel assemblies with 80-100 tonnes of ura­nium. The pressure vessel is of steel. Water under pressure is used both as coolant and moderator. The pressure vessel and the heat exchanger are surrounded by a concrete shield. In this reactor, bulk boiling water is prevented as the water is pressurized to about 150 atmospheres.

The hot water from the reactor flows to a heat exchanger (or steam generator) where its heat is transferred to the feed water to generate steam. The secondary cooling operates at a low pressure. The primary coolant then flows from the heat exchanger to the primary circulating pump which pumps it back to the reactor. The steam is condensed in the condenser and the condensate returns to heat exchanger forming a closed circuit. The primary circuit of a pressurized water reactor (PWR) contains a ‘pressurizer’.

This is simply a pressure vessel with an electric heating coil at the bottom and a water spray at the top. The top of the vessel is filled with steam at primary circuit pressure. When the primary circuit pressure decreases, the heating coil gets energized and boils the water to form steam resulting in increase in steam content in the vessel.

This results in the increase in pressure of the primary circuit. In case the steam pressure of the primary circuit becomes too high cold water is sprayed into the steam in the pressurizer. The steam is condensed and therefore primary circuit pressure is reduced. The steam generated is of rather poor quality, temperature around 250°C and pressure 42 kg/cm².

Its advantages are:

(i) Compactness,

(ii) Possibility of breed­ing plutonium,

(iii) Isolation of radioactive materials from the main steam system,

(iv) Cheap light water can be used as coolant-cum-moderator,

(v) High power density, and

(vi) The reactor responds to supply more power when the load in­creases. The positive power demand coefficient makes this almost automatic.

However, it suffers from the following drawbacks:

(i) Use of high pressure water system. So a strong pres­sure vessel is required which results in high cost.

(ii) Formation of low temperature (250° C) steam.

(iii) Use of expensive cladding material for prevention of corrosion.

(iv) High losses from heat exchanger.

(v) High power consumption by auxiliaries.

(vi) In comparison to other types, requirement of more elaborate safety devices.

(vii) These reactors cannot be re-fuelled while operating and for recharging the reactor is to be shut down for a couple of months. Also, there is difficulty in fuel element design and fabrication.

(viii) The thermodynamic efficiency of this plant is low (about 20%) due to low pressure in the secondary circuit.

The reactors installed at Rajasthan Atomic Power Sta­tion, Madras Atomic Power Station and Narora Atomic Power Project are of pressurized water reactor type. However they use heavy water as coolant and moderator. Such reactors are known as pressurized heavy water reactors (PHWR).

3. Gas Cooled Reactor:

This type of reactor employs a gas (CO₂ or helium) in place of water as the coolant and graphite as the moderator. A heat exchanger is necessarily required. Gas is circulated through the reactor core and the heat exchanger by means of a blower or a gas compressor. Even though gas is inferior to water from the point of view of heat transfer properties but it offers numerous advantages which are not available with water.

A large quantity of gas is required, due to poor heat transfer qualities, for circula­tion resulting in increased power consumption for auxilia­ries. Thus, advantage of high thermal efficiency is to a large extent lost and overall plant efficiency is low. Graphite, as moderator is less effective than water and would require a large volume core in such reactors, the heat removal by gas cooling will be better. The gas is circulated at a pressure of 14-28 kg/cm². The tubes in the heat exchanger through which water is circulated should have fins on their surface so as to improve the rate of heat transfer.

Its advantages are:

(i) Less severe corrosion problems.

(ii) Possibility of use of natural uranium as fuel.

(iii) Greater safety in comparison with water cooled re­actors.

(iv) Contamination problems are moderate.

(v) Low pressure coolant and relatively high reactor temperature.

Its drawbacks are:

(i) Relatively large size of reactor because of use of natural fuel and graphite moderator,

(ii) Extremely low power density,

(iii) Low steam pressure and temperature, and

(iv) Large energy consumption by gas blowers because of poor heat transfer characteristics of gases.

4. Heavy Water Cooled and Moderated (CANDU TYPE) Reactor:

This reactor was first developed by Canada and is, therefore, known as CANDU type reactor. The word CANDU stands for Canadian Deuterium Uranium. These reactors make use of heavy water, composed of the heavy hydrogen isotope, ₁H², as moderator to have maximum neutron economy and as coolant also. Such reactors are meant for those countries which do not have uranium enrichment facilities. Enrichment of uranium is costly affair and such reactors use natural uranium as fuel.

The primary and secondary circuits are similar to pressurized water reactor (PWR)—the coolant heavy water is circulated in the primary circuit and the steam is produced in the secondary circuit transferring the heat in the heat exchanger. Heavy hydrogen exists in nature in the ratio 1 : 6700 as compared to ordinary hydrogen and therefore, heavy water is very difficult and expensive to separate from ordinary water.

However, it is simpler to accomplish in comparison to enrichment of uranium. Hence in some designs, heavy water is used as moderator and light water is used in the secondary circuit. Control rods are not required in such reactors as the reactor control is achieved by varying the moderator level in the reactor. For rapid shut down purposes the moderator can be dumped through a very large area into a tank provided below the reactor.

The most important advantage of such a reactor is that the heavy water has a very low absorption cross section and it can be used as a moderator in natural uranium thermal reactors and, therefore, the fuel need not be enriched. Other advantages are simpler reactor control because of absence of control rods, high multiplication factor, low fuel consumption and much more effectiveness in slowing down neutrons because of moderator being at low temperature. It is worth-mentioning here that a major part of the equipment for this reactor can be manufactured in the shop and period required for site construction is also comparatively smaller.

Its main drawbacks are the heavy cost of heavy water, problems of leakage and very high standard design etc.

5. Liquid Metal Cooled Reactor:

This type of reactor has been developed to avoid difficulties faced in pressurization of water as in PWR and at the same time retaining the advantage of having high temperature. Metals in liquid state have good thermal conductivity and high temperature can be had at moderate pressure. However, handling of sodium introduces difficulties because of its activity in reactor core. It is therefore necessary to employ two heat transfer circuits so that radioactive sodium does not come in contact with the steam circuit.

Sodium is circulated through the reactor core and an intermediate heat exchanger where the heat from sodium (Na) is transferred to the NaK (an alloy of sodium and potassium) liquid metal which gives up heat in the heat exchanger to generate steam. Because of violent reaction of sodium with air and water the whole system should be leak tight.

Charg­ing and draining from one of the two loops should be done in an inert atmosphere to avoid the contact of Na or NaK with air. Sodium graphite reactor (SGR) uses slightly enriched uranium alloy or uranium carbide clad with stainless steel as fuel, graphite as moderator and liquid sodium as coolant.

NaK alloy has a lower melting point and therefore al­lows higher heat absorption. But potassium may react with graphite and also it has a higher neutron absorption cross section than sodium. Because of this, liquid sodium is used as a coolant in the primary heat transfer circuit. Again so­dium does not interact with stainless steel up to 600°C.

Its special features are:

(i) Elimination of pressure on reactor and primary circuit due to high boiling point of liquid metal,

(ii) Steam generation at high pressure and temperature,

(iii) Reduced corrosion problems,

(iv) High reactor temperature, and

(v) Reduced containment require­ments because of low coolant pressure.

The main disadvantages of sodium graphite reactor are:

(i) Relatively complex core,

(ii) Requirement of enriched fuel, and

(iii) Requirement of triple cycle cooling system with dual heat exchangers to minimize hazards etc.

6. Fast Breeder Reactor:

Such reactors are designed to produce more fissile material (Plutonium) than they consume (Thorium Th-232). A fast breeder reactor is a small vessel in which the required quantity (correspond­ing to critical mass) of enriched uranium or plutonium is kept without a moderator. The fissionable fuel core is surrounded by a blanket of fertile material (U-238 or Th- 232).

The fertile material (U-238 or Th-232) absorbs neu­trons produced by the fissioning of U-235 and produces fissile material Pu-239 or U-233 respectively. Two heat exchangers are used. The reactor core is cooled by liquid metal (sodium or potassium). In the second heat exchanger the coolant is again liquid sodium/potassium which transfers heat to feed water to generate steam.

This prevents the possibility of a sodium-water reac­tion with the radioactive sodium. 

In fast breeder reactors neutron shielding is provided by using boron, light water, oil or graphite. Gamma-ray shielding is accomplished by lead, concrete with added magnetite or barium etc.

The core of a fast reactor needs high enrichment (above 10% of fissile material). To reduce the fuel cost effect, it is imperative to employ high ratings. The core consists of 30% fuel, 50% coolant and 20% canning and structural material by volume.

The power density in a fast breeder reactor is considerably higher than in normal reactors. Therefore, liquid sodium, which is an efficient coolant and does not moderate neutrons, is used to take away heat produced in the core. The efficiency obtained with liquid sodium is about 42%, whereas with other coolants it is about 28%.

An important advantage of FBR technology is that it can use Thorium (as fertile material) which gets converted to U-233, a fissile isotope. This holds great promise for India as we have one of the world’s largest deposit of Thorium—about 450,000 tons in form of sand dunes in Kerala and along Gopalpur Chattarpur coast of Orissa.

Typical power densities (MW/m) in fission reactor cores are – Gas cooled 0.53; High temperature gas cooled 7.75; Heavy water 18.0; Boiling water 29.0; Pressurized water 54.75 and Fast breeder reactor 760.0.

Comparison of Thermal and Fast Breeder Reactors:

Thermal reactors have the following advantages and disadvantages as compared to fast breeder reactors:

Advantages:

1. Heat developed per unit volume of core or per unit area of fuel surface is less.

2. Ease of control.

3. Greater inherent safety.

Disadvantages:

1. Severely limited choice of fuel from the point of view of neutron economy when fuel used is uranium.

2. Much larger size and weight of reactor per unit power.

3. More fissile material consumption than could be automatically replaced.

Fast breeder reactors can convert more fertile material to fissile material and therefore, net fuel consumption is much less. As a matter of fact more fissile material could be produced than would be consumed by it (fast breeder reactor).