The first experiments to create a high-speed models of locomotives in the Soviet Union began in the 1930s. In 1934, at the Kolomna plant carried out preliminary designs of high-speed trains.The Russians want to copy the USA’s first Jet Train.
Don Wetzel, an engineer for the New York Central Railroad, was given the task in the mid-1960s of trying to make trains safer, less expensive and faster.
His solution: strap two jet engines to the roof of a locomotive and see what happens.
What happened was Wetzel created the first jet-powered train that even to this day is the fastest locomotive in America.
A turbojet train is a train powered by turbojet engines. Like a jet aircraft, but unlike a gas turbine locomotive, the train is propelled by the jet thrust of the engines, rather than by its wheels. Only a handful of jet-powered trains have been built, for experimental research in high-speed rail. Turbojet engines have been built with the engine incorporated into a railcar combining both propulsion and passenger accommodation rather than as separate locomotives hauling passenger coaches.
Turbojet engines are most efficient at high speeds and so they have been applied to high-speed passenger services, rather than freight.
Some time ago we had a few photos of a piece of technology called “Soviet Turbojet Train”. The projected speed for this out-of-the-sixties monster was planned to be up to 360 km/h, and it set a record of 250 km/h on the Soviet standard railway. The project was discarded afterwards, partly due to the very high fuel consumption of the jet engines compared to the engines of jet planes, and we thought the only train built was lost, but recently these guys discovered it rusting on the back ways of some railroad.
The first attempt to use turbojet engines on a railroad was made by the New York Central Railroad in 1966. Their railcar M-497 was able to reach speeds up to 184 miles per hour (296 km/h) – we will cover that next week
The Russian train maker Kalininsky formed the Speed Wagon Laboratory. Following the New Yorker’s example, the modified the chassis of one of their ER22 head engines to look more or less like a rough version of a Shinkansen, the Japanese bullet train which was already working in 1964 at 130 mph (210km/h).
They added two turbojet engines on the front as well: two turbojets from a Yakovlev YAK-40. Their first test was in 1971 on the line joining Golutvin with Ozery. They achieved a low 116mph (187km/h). However, they kept increasing the speed until they got up to 154mph (249km/h).
Like it’s American counterpart it never really went any further than that. Jet fuel costs, noise levels, and probably just the fact that this is plane old silly contributed to the closing of the programs in both countries..
A look inside the physics of achieving high-pressure water for waterjet cutting
The energy required for cutting materials is obtained by pressurizing water to ultra-high pressures and forming an intense cutting stream by focusing this high-speed water through a small, precious-stone orifice. There are two main steps involved in the waterjet cutting process.
The Electric Servo Pump generally pressurizes normal tap water at pressure levels above 50,000 psi; to produce the energy required for cutting.
Water is then focused through a small precious stone orifice to form an intense cutting stream. The stream moves at a velocity of up to 3 times the speed of sound, depending on how the water pressure is exerted.
The process is applicable to both water only and abrasive jets. For abrasive cutting applications, abrasive garnet is fed into the abrasive mixing chamber, which is part of the cutting head body, to produce a coherent and an extremely energetic abrasive jet stream.
To achieve these pressures, water is introduced into the unit by way of a booster pump and filter. This filtering process is very important as water must be clean before reaching ultra-high pressures in order to protect the high-pressure parts and provide a consistent cutting stream.
A water treatment system is sometimes needed to remove harmful minerals from the water. After being filtered, the water enters the high pressure cylinder where it is pressurized to the desired level.
The water is then carried to either an abrasive or straight-water cutting nozzle, depending on the application. The cutting nozzle can be stationary or integrated into motion equipment, which allows for intricate shapes and designs to be cut. Motion equipment can range from a simple cross-cutter to 2D systems and 3D machines as well as multiple axis robots. CAD/CAM software combined with CNC controllers translate drawings or commands into a digitally programmed path for the cutting head to follow.Cutting harder materials requires adding a fine mesh abrasive to the cutting stream. Various abrasive materials which can be used include olivine, garnet, and corundum with a particle size of between 50 to 120 mesh. When abrasive is required, TECHNI Waterjet provides an abrasive unit consisting primarily of an abrasive hopper, an abrasive feeder system, a pneumatically controlled on/off valve, and the abrasive cutting nozzle which contains the specialized mixing chamber.
The abrasive is first stored in the pressurized hopper and travels to a metering assembly, which controls the amount of particles fed to the nozzle. The abrasive is then introduced into the cutting stream in a special mixing chamber within the abrasive cutting head. Abrasive cutting allows harder materials to be cut at a faster rate by accelerating the erosion process. After the cut, residual energy from the cutting stream is dissipated in a catcher tank, which stores the kerf material and spent abrasive.
Relationship Between Increased Pressure and Cutting Speed
As pressure increases the power requirement increases proportionately and therefore with a given amount of available power the flow rate must be proportionately reduced, by using a smaller orifice, as shown in the commonly used formula P (Power) = p (pressure) x q (water flow rate).
For example, a 50% increase in pressure will require a 50% increase in power unless there is an equivalent reduction in flow rate (using a smaller orifice). Higher pressure gives an increase in cutting speed for a given amount of power, as higher pressures and lower volumes result in higher velocity of the water leaving the cutting head, which is a more efficient transfer of power to kinetic energy (the energy used in the cutting process).
This efficiency comes about because increasing velocity is a more efficient way of increasing the kinetic energy stored in the particles of abrasive hitting the work-piece. This is illustrated through another commonly used formula E=M V², where by increasing the velocity has a squared affect on the kinetic energy, compared to increasing the mass which has a linear affect. Therefore, in theory if we increase the pressure by 50%, but decrease the volume by 33% we use the same amount of power but get an increase in the velocity of 50%, which has the effect of increasing the kinetic energy by 48.5%, as illustrated in the formula E=0.666 (33% reduction in water mass) x 1.5² (50% increase in velocity) therefore E = 1.485 (48.5% increase in kinetic energy). However, this illustration is only relevant for Water Only cutting, as the mass of the abrasive has not yet been taken into account.
Abrasive cutting dramatically increases the cutting capabilities of a waterjet by accelerating the abrasive particles at the work piece where each particle takes out a small gouge of the work piece material during impact. If all the abrasive particles were to hit the work piece in the same condition, but at the higher velocity, the same equation as above would be true. However, the major factor that affects what actually happens is that the abrasive particles get smashed to a very fine powder when hit by the high velocity water stream during the initial introduction of the abrasive to the stream, and more gets destroyed throughout the focussing tube 1 The intensity of the disintegration of the abrasive particles depends on the water pressure. The result is that at 60,000psi, only about 45% of the abrasive material reaches the work piece in an affective cutting condition. This % drops to about 22% (or less depending on the quality of abrasive) at 90,000 psi. The net result is that there is therefore only a very small net increase in cutting speed when pressure is increased, for an equivalent amount of power.This can be illustrated in simplified terms for 60,000psi as E=0.45 (45% effective garnet) x 1² therefore E=0.45, and 90,000psi as E=0.22 (22% effective garnet) x 1.5² (50% increase in velocity) therefore E=0.49, or a 9% increase in cutting speed for a 50% increase in pressure.
The Cost of Higher Pressures
Another important consideration before deciding to increase pressure is the significant increase in the capital cost, maintenance cost, consumable cost and increased machine downtime.
Pressure (also known as force or load) has a non-linear relationship with fatigue-related wear, and for many mechanical machine components, it has a cubed (x³) relationship. For example, the ISO formula for calculating bearing wear is
L = (C/P)3 L = life C= rated load P = actual load
That means that a 50% increase in pressure will reduce the design life of many mechanical components by about 70% or adversely by reducing pressure by 33%, say from 90,000psi to 60,000psi, the life of many components will increase by 330%.
As a result, in order to make pumps and cutting heads that last for reasonable amounts of time at extreme pressures like 90,000psi, manufacturers are forced to use very expensive exotic materials, because metal fatigue becomes the dominant failure mechanism.. The cost of components and consumables that experience high pressures over 66,000psi (such as dynamic and static seals, check valves, tubing and high- and low-pressure cylinders) are therefore typically 50-300% higher than standard waterjet components.
The other factor in the pricing of such components is competition, as only a few manufacturers are currently capable of producing those parts that are resistant to wear and failure at high pressures. As a consequence, the competition to bring down prices doesn’t yet exist.
For instance, a standard focussing tube rated to 60,000psi sells for approx. $100, while a 90,000psi rated tube sells for around $300. Moreover, even with the more exotic and expensive components designed for 90,000psi, their life remains well below that of traditional waterjet parts operating with up to 60,000psi. This means increased down time and higher maintenance labour costs, on top of the higher component prices and more frequent part replacements.
With that you can understand that more cc does not mean more power. So, that brings to the most asked question: Does more cc mean more fuel consumption?
From what I've read, it is generally true that a vehicle with more displacement will have more fuel consumption.
However, there are very many other factors that affect fuel efficiency.
Therefore, by minimizing engine displacement, you will not ensure increased fuel efficiency.
2. What is meant by 'wheelbase' and 'ground clearance' in a car?
Wheelbase:
In a car, there are two rods used to connect the center of the wheels, one on the front and another on the rear.
The distance between these two rods or axles of a vehicle is known as its wheelbase.
This term is generally checked while buying a car to see how much large the cabin is. The longer the wheelbase, the more the interior room in the car's cabin.
Ground Clearance:
It means the distance between the ground (the point where the tire meets the ground) and the under side of the chassis i.e. any parts that aren't designed to touch the ground. The manufacturers generally mention this distance in millimetres or inches in the list of specifications.
High Ground Clearance Vs. Low Ground Clearance:
More the ground clearance, more is the vehicle capable of moving on off-road, bumpy, rough terrains. For this reason, you will generally find that SUVs are designed with high ground clearance.
If the vehicles has low ground clearance will have low center of gravity and that leads to better handling and performance.
A balance between a high and low ground clearance is needed and you'll find this being achieved in the most executive sedans.
3.What does air-cooled engine mean?
As is apparent from the term we are looking at, Air-cooled engines have air circulating over the hot parts of the engine to cool it. Now I can't put a diagrammatic explanation here, but it is pretty clear even with just the theory.
Most cars we use today have modern internal combustion engines.
A great percentage of the heat generated through these engines is released through the exhaust.
The remaining is handled traditionally by using a liquid coolant that is passed through a closed circuit over the cylinder head and engine block.
The liquid coolant absorbs heat and when it reaches the heat exchanger or radiator of the car, it released the heat into air.
Now you may feel that isn't the engine being ultimately cooled by air. Well yes, but because a liquid-coolant circuit was used, this system is called water-cooled engine.
In contrast, the air-cooled engines have the heat generated released directly into the air. Natural air flow plays a big part in this.
4.What is the difference between Multi Utility Vehicle (MUV) and Sports Utility Vehicle (SUV)?
SUV is a term used to denote any vehicle that looks like a station wagon.
They are equipped with four-wheel drive or all-wheel drive.
Their design aims to display superior off-road and towing capabilities and bigger seating capacity.
Example: Mahindra Scorpio, Ford Ecosport, Renault Duster, Toyota Fortuner etc.
MUV is a type of vehicle designed in a shape of van.
They typically allow easy conversion between multiple combinations of passenger and luggage capacity.
Example: Toyota Innova, Maruti Ertiga, Renaulta Lodgy, Honda Mobilio etc.
5. What is the difference between Automatic transmission and Manual transmission?
Automatic transmission have only few select options, like forward, neutral and reverse. Where manual transmission will have complete gear selector.
For buyers, Automatic means a vehicle will select a appropriate Gear by itself as 1st, 2nd or .... as per the vehicle speed and load. where in Manual a driver has to decide and put vehicle in correct gear to move in desired form.
Automatic may have solenoid gear selector or a complex mechanism of overdrive and clutches which we see in luxurious vehicles only.
6. Front and Rear suspension - What do they mean and why are they needed?
First why suspension, between you vehicle body and wheels suspension are the main medium which holds vehicle and make moving when moving wheels. second suspension is for absorbing sudden shocks from round terrain. Each wheel have separate suspension for independent work( minds its for cars only, trucks may have different arrangements)
in every cars there are two axles front and rear. each axles will get two suspension of same type on both the wheels.
For buyers, Front suspension has some difference bcoz front suspension has to modify and adjust as per the turning on front wheel in corners. whereas rear wheels has to Go straight only.
7. What is the difference between 4-cylinder engine and 3-cylinder engine?
4 or 3 cylinder engine - it implies number of cylinder and piston you have in your engine. More number of cylinder means more CC and more power. and also it increases size of the engine.
8. What is drivetrain and powertrain and why is it important?
Powertrain
The powertrain is composed of everything that makes the vehicle move. These components include the engine which generates the initial energy, the transmission that distributes it and produces torque and all the other components of the drivetrain that help to propel the body forward. It can be expressed in terms of a mathematical equation for simpler understanding:
Powertrain = [Engine] + [Drivetrain]
The output from the power sources are controlled by a transmission system and the driveline to deliver torque to the wheels. The circular motion of the crankshaft is transmitted to the rear wheels through the gearbox, clutch, universal joints, drive shafts or propeller shafts, the differential and the axles connected to the wheels.
The application of engine power to the driving wheels through the collaborative effort of each of these components is called the power transmission. All wheel drive vehicles have two sets of these components to distribute the power almost equally to the front and the rear.
Drivetrain
The drivetrain is the part of a motorized carriage that connects the engine and transmission to the wheel axles through a number of other components. Drivetrain consists of all components after the transmission.
9. What is the difference between All-wheel drive and four-wheel drive?
All wheel drive - means the power as explained in power goes from engine to all wheels of a vehicle basically the term uses for multi axle heavy duty trucks or trailer tractors.
Four Wheel Drive- as it says power goes to 4 wheel drive only.
4×4 (also, four-wheel drive and 4WD)
Reflecting two axles with both wheels on each capable of being powered.
6×6 (also, six-wheel drive and 6WD)
Reflecting three axles with both wheels on each capable of being powered.
8×8 (also, eight-wheel drive and 8WD)
Reflecting four axles with both wheels on each capable of being powered.
The cutting tool is one of the most important things to consider in the machining of metal in the lathe. In order to machine metal accurately and efficiently it is important that the cutter bit have a keen cutting edge, ground with the correct clearance, rake, etc., for the particular kind of metal being machined, and that the cutter bit be set at the correct height.
In this booklet the latest shop practice for grinding various types of lathe tool cutter bits is outlined.
The cutter bit is that part of the lathe tool which cuts the metal that must be removed to bring the work to the desired size and shape. The cutter bit is usually made of high speed steel and held in a lathe tool holder
A U engine is a piston engine made up of two separate straight engines (complete with separate crankshafts) joined by gears or chains. It is similar to the H engine which couples two flat engines.
The design is also sometimes described as a "twin bank" or "double
bank" engine, although these terms are sometimes used also to describe V engines.
This configuration is uncommon, as it is heavier than a V design. The
main interest in this design is its ability to share common parts with
straight engines. However, V engines with offset banks can also share
straight engine parts (except for the crankshaft), and this is therefore
a far more common design today when both engine forms are produced from
the same basic design.
A nuclear power plant or nuclear power station is a thermal power station in which the heat source is a nuclear reactor. As is typical in all conventional thermal power stations the heat is used to generate steam which drives a steam turbine connected to an electric generator which produces electricity.
Most nuclear electricity is generated using just two kinds of reactors which were developed in the 1950s and improved since.
New designs are coming forward and some are in operation as
the first generation reactors come to the end of their operating lives.
Over 11% of the world's electricity is produced from nuclear energy, more than from all sources worldwide in 1960.
This paper is about the main conventional types of nuclear reactor. For more advanced types, see Advanced Reactors and Small Reactors papers, and also Generation IV reactors.
A nuclear reactor produces and controls the release of energy from
splitting the atoms of certain elements. In a nuclear power reactor, the
energy released is used as heat to make steam to generate electricity.
(In a research reactor the main purpose is to utilise the actual
neutrons produced in the core. In most naval reactors, steam drives a
turbine directly for propulsion.)
The principles for using nuclear power to produce electricity are the
same for most types of reactor. The energy released from continuous
fission of the atoms of the fuel is harnessed as heat in either a gas or
water, and is used to produce steam. The steam is used to drive the
turbines which produce electricity (as in most fossil fuel plants).
The world's first nuclear reactors operated naturally in a uranium
deposit about two billion years ago. These were in rich uranium
orebodies and moderated by percolating rainwater. The 17 known at Oklo
in west Africa, each less than 100 kW thermal, together consumed about
six tonnes of that uranium. It is assumed that these were not unique
worldwide.
Today, reactors derived from designs originally developed for
propelling submarines and large naval ships generate about 85% of the
world's nuclear electricity. The main design is the pressurised water
reactor (PWR) which has water at over 300°C under pressure in its
primary cooling/heat transfer circuit, and generates steam in a
secondary circuit. The less numerous boiling water reactor (BWR) makes
steam in the primary circuit above the reactor core, at similar
temperatures and pressure. Both types use water as both coolant and
moderator, to slow neutrons. Since water normally boils at 100°C, they
have robust steel pressure vessels or tubes to enable the higher
operating temperature. (Another type uses heavy water, with deuterium
atoms, as moderator. Hence the term ‘light water’ is used to
differentiate.)
Components of a nuclear reactor
There are several components common to most types of reactors:
Fuel. Uranium is the basic fuel. Usually pellets of uranium oxide (UO2) are arranged in tubes to form fuel rods. The rods are arranged into fuel assemblies in the reactor core.*
* In a new reactor with new fuel a neutron source
is needed to get the reaction going. Usually this is beryllium mixed
with polonium, radium or other alpha-emitter. Alpha particles from the
decay cause a release of neutrons from the beryllium as it turns to
carbon-12. Restarting a reactor with some used fuel may not require
this, as there may be enough neutrons to achieve criticality when
control rods are removed.
Moderator. Material in the core which slows down the
neutrons released from fission so that they cause more fission. It is
usually water, but may be heavy water or graphite.
Control rods. These are made with neutron-absorbing
material such as cadmium, hafnium or boron, and are inserted or
withdrawn from the core to control the rate of reaction, or to halt
it.* In some PWR reactors, special control rods are used to enable the
core to sustain a low level of power efficiently. (Secondary control
systems involve other neutron absorbers, usually boron in the coolant –
its concentration can be adjusted over time as the fuel burns up.)
* In fission, most of the neutrons are released promptly,
but some are delayed. These are crucial in enabling a chain reacting
system (or reactor) to be controllable and to be able to be held
precisely critical.
Coolant. A fluid circulating through the core so as
to transfer the heat from it. In light water reactors the water
moderator functions also as primary coolant. Except in BWRs, there is
secondary coolant circuit where the water becomes steam. (See also later
section on primary coolant characteristics)
Pressure vessel or pressure tubes. Usually a robust
steel vessel containing the reactor core and moderator/coolant, but it
may be a series of tubes holding the fuel and conveying the coolant
through the surrounding moderator.
Steam generator. Part of the cooling system of
pressurised water reactors (PWR & PHWR) where the high-pressure
primary coolant bringing heat from the reactor is used to make steam for
the turbine, in a secondary circuit. Essentially a heat exchanger like a
motor car radiator*. Reactors have up to six 'loops', each with a steam
generator. Since 1980 over 110 PWR reactors have had their steam
generators replaced after 20-30 years service, 57 of these in USA.
* These are large heat exchangers for
transferring heat from one fluid to another – here from high-pressure
primary circuit in PWR to secondary circuit where water turns to steam.
Each structure weighs up to 800 tonnes and contains from 300 to 16,000
tubes about 2 cm diameter for the primary coolant, which is radioactive
due to nitrogen-16 (N-16, formed by neutron bombardment of oxygen, with
half-life of 7 seconds). The secondary water must flow through the
support structures for the tubes. The whole thing needs to be designed
so that the tubes don't vibrate and fret, operated so that deposits do
not build up to impede the flow, and maintained chemically to avoid
corrosion. Tubes which fail and leak are plugged, and surplus capacity
is designed to allow for this. Leaks can be detected by monitoring N-16
levels in the steam as it leaves the steam generator.
Containment. The structure around the reactor and
associated steam generators which is designed to protect it from outside
intrusion and to protect those outside from the effects of radiation in
case of any serious malfunction inside. It is typically a metre-thick
concrete and steel structure.
Newer Russian and some other reactors install core melt localisation
devices or 'core catchers' under the pressure vessel to catch any melted
core material in the event of a major accident.
There are several different types of reactors as indicated in the following table.
Nuclear power plants in commercial operation
Reactor type
Main Countries
Number
GWe
Fuel
Coolant
Moderator
Pressurised water reactor (PWR)
US, France, Japan, Russia, China
277
257
enriched UO2
water
water
Boiling water reactor (BWR)
US, Japan, Sweden
80
75
enriched UO2
water
water
Pressurised heavy water reactor (PHWR)
Canada, India
49
25
natural UO2
heavy water
heavy water
Gas-cooled reactor (AGR & Magnox)
UK
15
8
natural U (metal), enriched UO2
CO2
graphite
Light water graphite reactor (RBMK & EGP)
Russia
11 + 4
10.2
enriched UO2
water
graphite
Fast neutron reactor (FBR)
Russia
2
0.6
PuO2 and UO2
liquid sodium
none
TOTAL
438
376
IAEA data, end of 2014. GWe = capacity in thousands of megawatts (gross)
Source: Nuclear Engineering International Handbook 2011, updated to 1/1/12
For reactors under construction: see paper Plans for New Reactors Worldwide.
Fuelling a nuclear power reactor
Most reactors need to be shut down for refuelling, so that the
pressure vessel can be opened up. In this case refuelling is at
intervals of 1-2 years, when a quarter to a third of the fuel assemblies
are replaced with fresh ones. The CANDU and RBMK types have pressure
tubes (rather than a pressure vessel enclosing the reactor core) and can
be refuelled under load by disconnecting individual pressure tubes.
If graphite or heavy water is used as moderator, it is possible to
run a power reactor on natural instead of enriched uranium. Natural
uranium has the same elemental composition as when it was mined (0.7%
U-235, over 99.2% U-238), enriched uranium has had the proportion of the
fissile isotope (U-235) increased by a process called enrichment,
commonly to 3.5 - 5.0%. In this case the moderator can be ordinary
water, and such reactors are collectively called light water reactors.
Because the light water absorbs neutrons as well as slowing them, it is
less efficient as a moderator than heavy water or graphite.
During operation, some of the U-238 is changed to plutonium, and Pu-239 ends up providing about one third of the energy from the fuel.
In most reactors the fuel is ceramic uranium oxide (UO2
with a melting point of 2800°C) and most is enriched. The fuel pellets
(usually about 1 cm diameter and 1.5 cm long) are typically arranged in a
long zirconium alloy (zircaloy) tube to form a fuel rod, the zirconium
being hard, corrosion-resistant and transparent to neutrons.* Numerous
rods form a fuel assembly, which is an open lattice and can be lifted
into and out of the reactor core. In the most common reactors these are
about 4 metres long. A BWR fuel assembly may be about 320 kg, a PWR one
655 kg, in which case they hold 183 kg uranium and 460 kgU respectively.
In both, about 100 kg of zircaloy is involved.
*Zirconium is an important mineral for nuclear power,
where it finds its main use. It is therefore subject to controls on
trading. It is normally contaminated with hafnium, a neutron absorber,
so very pure 'nuclear grade' Zr is used to make the zircaloy, which is
about 98% Zr plus about 1.5% tin, also iron, chromium and sometimes
nickel to enhance its strength.
Burnable poisons are often used in fuel or coolant to even out the
performance of the reactor over time from fresh fuel being loaded to
refuelling. These are neutron absorbers which decay under neutron
exposure, compensating for the progressive build up of neutron absorbers
in the fuel as it is burned. The best known is gadolinium, which is a
vital ingredient of fuel in naval reactors where installing fresh fuel
is very inconvenient, so reactors are designed to run more than a decade
between refuellings. Gadolinium is
incorporated in the ceramic fuel pellets. An alternative is zirconium
diboride integral fuel burnable absorber (IFBA) as a thin coating on
normal pellets.
Gadolinium, mostly at up to 3g oxide per kilogram of fuel, requires
slightly higher fuel enrichment to compensate for it, and also after
burn-up of about 17 GWd/t it retains about 4% of its absorbtive effect
and does not decrease further. The ZrB2 IFBA burns away more
steadily and completely, and has no impact on fuel pellet properties. It
is now used in most US reactors and a few in Asia. China has the
technology for AP1000 reactors.
The power rating of a nuclear power reactor
Nuclear power plant reactor power outputs are quoted in three ways:
Thermal MWt, which depends on the design of the actual nuclear
reactor itself, and relates to the quantity and quality of the steam it
produces.
Gross electrical MWe indicates the power produced by the attached
steam turbine and generator, and also takes into account the ambient
temperature for the condenser circuit (cooler means more electric power,
warmer means less). Rated gross power assumes certain conditions with
both.
Net electrical MWe, which is the power available to be sent out
from the plant to the grid, after deducting the electrical power needed
to run the reactor (cooling and feed-water pumps, etc.) and the rest of
the plant.*
* Net electrical MWe and gross MWe vary
slightly from summer to winter, so normally the lower summer figure, or
an average figure, is used. If the summer figure is quoted plants may
show a capacity factor greater than 100% in cooler times. Watts Bar PWR
in Tennessee is reported to run at about 1125 MWe in summer and about
1165 MWe net in winter, due to different condenser cooling water
temperatures. Some design options, such as powering the main large
feed-water pumps with electric motors (as in EPR) rather than steam
turbines (taking steam before it gets to the main turbine-generator),
explains some gross to net differences between different reactor types.
The EPR has a relatively large drop from gross to net MWe for this
reason.
The relationship between these is expressed in two ways:
Thermal efficiency %, the ratio of gross MWe to thermal MW. This
relates to the difference in temperature between the steam from the
reactor and the cooling water. It is often 33-37%.
Net efficiency %, the ratio of net MWe achieved to thermal MW. This is a little lower, and allows for plant usage.
In WNA papers and figures and WNN items, generally net MWe is used
for operating plants, and gross MWe for those under construction or
planned/proposed.
Pressurised water reactor (PWR)
This is the most common type, with over 230 in use for power
generation and several hundred more employed for naval propulsion. The
design of PWRs originated as a submarine power plant.
PWRs use ordinary water as both coolant and moderator. The design is
distinguished by having a primary cooling circuit which flows through
the core of the reactor under very high pressure, and a secondary
circuit in which steam is generated to drive the turbine. In Russia
these are known as VVER types – water-moderated and -cooled.
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.
Water in the reactor core reaches about 325°C, hence it must be kept
under about 150 times atmospheric pressure to prevent it boiling.
Pressure is maintained by steam in a pressuriser (see diagram). In the
primary cooling circuit the water is also the moderator, and if any of
it turned to steam the fission reaction would slow down. This negative
feedback effect is one of the safety features of the type. The secondary
shutdown system involves adding boron to the primary circuit.
The secondary circuit is under less pressure and the water here boils
in the heat exchangers which are thus steam generators. The steam
drives the turbine to produce electricity, and is then condensed and
returned to the heat exchangers in contact with the primary circuit.
Boiling water reactor (BWR)
This design has many similarities to the PWR, except that there is
only a single circuit in which the water is at lower pressure (about 75
times atmospheric pressure) so that it boils in the core at about 285°C.
The reactor is designed to operate with 12-15% of the water in the top
part of the core as steam, and hence with less moderating effect and
thus efficiency there. BWR units can operate in load-following mode
more readily then PWRs.
The steam passes through drier plates (steam separators) above the
core and then directly to the turbines, which are thus part of the
reactor circuit. Since the water around the core of a reactor is always
contaminated with traces of radionuclides, it means that the turbine
must be shielded and radiological protection provided during
maintenance. The cost of this tends to balance the savings due to the
simpler design. Most of the radioactivity in the water is very
short-lived*, so the turbine hall can be entered soon after the reactor
is shut down.
* mostly N-16, with a 7 second half-life
A BWR fuel assembly comprises 90-100 fuel rods, and there are up to
750 assemblies in a reactor core, holding up to 140 tonnes of uranium.
The secondary control system involves restricting water flow through the
core so that more steam in the top part reduces moderation.
Pressurised heavy water reactor (PHWR)
The PHWR reactor design has been developed since the 1950s in Canada
as the CANDU, and from 1980s also in India. PHWRs generally use natural
uranium (0.7% U-235) oxide as fuel, hence needs a more efficient
moderator, in this case heavy water (D2O).** The PHWR
produces more energy per kilogram of mined uranium than other
designs, but also produces a much larger amount of used fuel per unit
output. ** with the CANDU system, the moderator is enriched (i.e. water) rather than the fuel – a cost trade-off. The moderator is in a large tank called a calandria, penetrated by
several hundred horizontal pressure tubes which form channels for the
fuel, cooled by a flow of heavy water under high pressure in the primary
cooling circuit, reaching 290°C. As in the PWR, the primary coolant
generates steam in a secondary circuit to drive the turbines. The
pressure tube design means that the reactor can be refuelled
progressively without shutting down, by isolating individual pressure
tubes from the cooling circuit. It is also less costly to build than
designs with a large pressure vessel, but the tubes have not proved as
durable.
A CANDU fuel assembly consists of a bundle of 37 half metre long fuel
rods (ceramic fuel pellets in zircaloy tubes) plus a support structure,
with 12 bundles lying end to end in a fuel channel. Control rods
penetrate the calandria vertically, and a secondary shutdown system
involves adding gadolinium to the moderator. The heavy water moderator
circulating through the body of the calandria vessel also yields some
heat (though this circuit is not shown on the diagram above).
Newer PHWR designs such as the Advanced Candu Reactor (ACR) have light water cooling and slightly-enriched fuel.
CANDU reactors can accept a variety of fuels. They may be run on
recycled uranium from reprocessing LWR used fuel, or a blend of this and
depleted uranium left over from enrichment plants. About 4000 MWe of
PWR might then fuel 1000 MWe of CANDU capacity, with addition of
depleted uranium. Thorium may also be used in fuel.
Advanced gas-cooled reactor (AGR)
These are the second generation of British gas-cooled reactors, using
graphite moderator and carbon dioxide as primary coolant. The fuel is
uranium oxide pellets, enriched to 2.5-3.5%, in stainless steel tubes.
The carbon dioxide circulates through the core, reaching 650°C and then
past steam generator tubes outside it, but still inside the concrete and
steel pressure vessel (hence 'integral' design). Control rods penetrate
the moderator and a secondary shutdown system involves injecting
nitrogen to the coolant.
The AGR was developed from the Magnox reactor, also graphite moderated and CO2
cooled, and one of these is still operating in UK to late 2014. They
use natural uranium fuel in metal form. Secondary coolant is water.
Light water graphite-moderated reactor (RBMK)
This is a Soviet design, developed from plutonium production
reactors. It employs long (7 metre) vertical pressure tubes running
through graphite moderator, and is cooled by water, which is allowed to
boil in the core at 290°C, much as in a BWR. Fuel is low-enriched
uranium oxide made up into fuel assemblies 3.5 metres long. With
moderation largely due to the fixed graphite, excess boiling simply
reduces the cooling and neutron absorbtion without inhibiting the
fission reaction, and a positive feedback problem can arise, which is
why they have never been built outside the Soviet Union. See appendix on
RBMK Reactors for more detail.
Advanced reactors
Several generations of reactors are commonly distinguished.
Generation I reactors were developed in 1950-60s and only one is still
running today. They mostly used natural uranium fuel and used graphite
as moderator. Generation II reactors are typified by the present US
fleet and most in operation elsewhere. They typically use enriched
uranium fuel and are mostly cooled and moderated by water. Generation
III are the Advanced Reactors evolved from these, the first few of which
are in operation in Japan and others are under construction and ready
to be ordered. They are developments of the second generation with
enhanced safety. There is no clear distinction Gen II to Gen III.
Generation IV designs are still on the drawing board and will not be
operational before 2020 at the earliest, probably later. They will tend
to have closed fuel cycles and burn the long-lived actinides now forming
part of spent fuel, so that fission products are the only high-level
waste. Of seven designs under development, 4 or 5 will be fast neutron
reactors. Four will use fluoride or liquid metal coolants, hence operate
at low pressure. Two will be gas-cooled. Most will run at much higher
temperatures than today’s water-cooled reactors. See Generation IV Reactors paper.
More than a dozen (Generation III) advanced reactor
designs are in various stages of development. Some are evolutionary
from the PWR, BWR and CANDU designs above, some are more radical
departures. The former include the Advanced Boiling Water Reactor, a few
of which are now operating with others under construction. The
best-known radical new design has the fuel as large 'pebbles' and uses
helium as coolant, at very high temperature, possibly to drive a turbine
directly.
Considering the closed fuel cycle, Generation 1-3 reactors recycle
plutonium (and possibly uranium), while Generation IV are expected to
have full actinide recycle.
Fast neutron reactors (FNR)
Some reactors (only one in commercial service) do not have a
moderator and utilise fast neutrons, generating power from plutonium
while making more of it from the U-238 isotope in or around the fuel.
While they get more than 60 times as much energy from the original
uranium compared with the normal reactors, they are expensive to build.
Further development of them is likely in the next decade, and the main
designs expected to be built in two decades are FNRs. If they are
configured to produce more fissile material (plutonium) than they
consume they are called Fast Breeder Reactors (FBR). See also Fast Neutron Reactors and Small Reactors papers.
Floating nuclear power plants
Apart from over 200 nuclear reactors powering various kinds of ships,
Rosatom in Russia has set up a subsidiary to supply floating nuclear
power plants ranging in size from 70 to 600 MWe. These will be mounted
in pairs on a large barge, which will be permanently moored where it is
needed to supply power and possibly some desalination to a shore
settlement or industrial complex. The first has two 40 MWe reactors
based on those in icebreakers and will operate at a remote site in
Siberia. Electricity cost is expected to be much lower than from present
alternatives.
The Russian KLT-40S is a reactor well proven in icebreakers and now
proposed for wider use in desalination and, on barges, for remote area
power supply. Here a 150 MWt unit produces 35 MWe (gross) as well as up
to 35 MW of heat for desalination or district heating. These are
designed to run 3-4 years between refuelling and it is envisaged that
they will be operated in pairs to allow for outages, with on-board
refuelling capability and used fuel storage. At the end of a 12-year
operating cycle the whole plant is taken to a central facility for
2-year overhaul and removal of used fuel, before being returned to
service. Two units will be mounted on a 21,000 tonne barge. A larger
Russian factory-built and barge-mounted reactor is the VBER-150, of 350
MW thermal, 110 MWe. The larger VBER-300 PWR is a 325 MWe unit,
originally envisaged in pairs as a floating nuclear power plant,
displacing 49,000 tonnes. As a cogeneration plant it is rated at 200 MWe
and 1900 GJ/hr. See also Nuclear Power in Russia paper.
Lifetime of nuclear reactors
Most of today's nuclear plants which were originally designed for 30
or 40-year operating lives. However, with major investments in systems,
structures and components lives can be extended, and in several
countries there are active programs to extend operating lives. In the
USA most of the more than one hundred reactors are expected to be
granted licence extensions from 40 to 60 years. This justifies
significant capital expenditure in upgrading systems and components,
including building in extra performance margins.
Some components simply wear out, corrode or degrade to a low level of
efficiency. These need to be replaced. Steam generators are the most
prominent and expensive of these, and many have been replaced after
about 30 years where the reactor otherwise has the prospect of running
for 60 years. This is essentially an economic decision. Lesser
components are more straightforward to replace as they age. In Candu
reactors, pressure tube replacement has been undertaken on some plants
after about 30 years operation.
A second issue is that of obsolescence. For instance, older reactors
have analogue instrument and control systems. Thirdly, the properties of
materials may degrade with age, particularly with heat and neutron
irradiation. In respect to all these aspects, investment is needed to
maintain reliability and safety. Also, periodic safety reviews are
undertaken on older plants in line with international safety conventions
and principles to ensure that safety margins are maintained.
Another important issue is knowledge management (KM) over the full
lifecycle from design, through construction and operation to
decommissioning for reactors and other facilities. This may span a
century and involve several countries, and involve a succession of
companies. The plant lifespan will cover several generations of
engineers. Data needs to be transferable across several generations of
software and IT hardware, as well as being shared with other operators
of similar plants.* Significant modifications may be made to the design
over the life of the plant, so original documentation is not sufficient,
and loss of design base knowledge can have huge implications (eg
Pickering A and Bruce A in Ontario). Knowledge management is often a
shared responsibility and is essential for effective decision-making and
the achievement of plant safety and economics.
* ISO15926 covers portability and interoperability for lifecycle open data standard. Also EPRI in 2013 published Advanced Nuclear Technology: New Nuclear Power Plant Information Handover Guide.
See also section on Ageing, in Safety of Nuclear Power Reactors paper.
Load-following capability
Nuclear power plants are essentially base-load generators, ideally
running continuously at high capacity. This is because their power
output cannot efficiently be ramped up and down on a daily and weekly
basis, and in this respect they are similar to most coal-fired plants.
(It is also uneconomic to run them at less than full capacity, since
they are expensive to build but cheap to run.) However, in some
situations it is necessary to vary the output according to daily and
weekly load cycles on a regular basis, for instance in France, where
there is a very high reliance on nuclear power.
BWRs can be made to follow loads reasonably easily without burning
the core unevenly, by changing the coolant flow rate. Load following is
not as readily achieved in a PWR, but especially in France since 1981,
so-called 'grey' control rods are used. The ability of a PWR to run at
less than full power for much of the time depends on whether it is in
the early part of its 18 to 24-month refuelling cycle or late in it, and
whether it is designed with special control rods which diminish power
levels throughout the core without shutting it down. Thus, though the
ability on any individual PWR reactor to run on a sustained basis at low
power decreases markedly as it progresses through the refueling cycle,
there is considerable scope for running a fleet of reactors in
load-following mode. European Utility Requirements (EUR) since 2001
specify that new reactor designs must be capable of load-following
between 50 and 100% of capacity with a rate of change of electric output
of 3-5% per minute. The economic consequences are mainly due to
diminished load factor of a capital-intensive plant. Further information
in the Nuclear Power in France paper and the 2011 Nuclear Energy Agency report, Technical and Economic Aspects of Load Following with Nuclear Power Plants.
As fast neutron reactors become established in future years, their ability to load-follow will be a benefit.
Primary coolants
The advent of some of the designs mentioned above provides
opportunity to review the various primary heat transfer fluids used in
nuclear reactors. There is a wide variety – gas, water, light metal,
heavy metal and salt:
Water or heavy water must be maintained at very high
pressure (1000-2200 psi, 7-15 MPa, 150 atmospheres) to enable it to
function well above 100°C, up to 345°C, as in present reactors. This has
a major influence on reactor engineering. However, supercritical water
around 25 MPa can give 45% thermal efficiency – as at some fossil-fuel
power plants today with outlet temperatures of 600°C, and at ultra
supercritical levels (30+ MPa) 50% may be attained.
Water cooling of steam condensers is fairly standard in all power
plants, because it works very well, it is relatively inexpensive, and
there is a huge experience base. Water is a lot more effective than air
for removing heat, though its thermal conductivity is less than liquid
alternatives.
Helium must be used at similar pressure (1000-2000
psi, 7-14 MPa) to maintain sufficient density for efficient operation.
Again, there are engineering implications, but it can be used in the
Brayton cycle to drive a turbine directly.
Carbon dioxide was used in early British reactors,
and their current AGRs which operate at much higher temperatures than
light water reactors. It is denser than helium and thus likely to give
better thermal conversion efficiency. It also leaks less readily than
helium. There is now interest in supercritical CO2 for the Brayton cycle.
Sodium, as normally used in fast neutron reactors at
around 550ºC, melts at 98°C and boils at 883°C at atmospheric pressure,
so despite the need to keep it dry the engineering required to contain
it is relatively modest. It has high thermal conductivity. However,
normally water/steam is used in the secondary circuit to drive a turbine
(Rankine cycle) at lower thermal efficiency than the Brayton cycle. In
some designs sodium is in a secondary circuit to steam generators.
Sodium does not corrode the metals used in the fuel cladding or primary
circuit, nor the fuel itself if there is cladding damage, but it is very
reactive generally. In particular it reacts exothermically with water
or steam to liberate hydrogen. It burns in air, but much less
vigorously. Sodium has a low neutron capture cross section, but it is
enough for some Na-23 to become Na-24, which is a beta-emitter and very
gamma-active with 15-hour half-life, so some shielding is required. If a
reactor needs to be shut down frequently, NaK eutectic which is liquid
at room temperature (about 13°C) may be used as coolant, but the
potassium is pyrophoric, which increases the hazard.
Lead or lead-bismuth eutectic in fast neutron
reactors are capable of higher temperature operation at atmospheric
pressure. They are transparent to neutrons, aiding efficiency due to
greater spacing between fuel pins which then allows coolant flow by
convection for decay heat removal, and since they do not react with
water the heat exchanger interface is safer. They do not burn when
exposed to air. However, they are corrosive of fuel cladding and steels,
which originally limited temperatures to 550°C. With today's materials
650°C can be reached, and in future 800°C is envisaged with the second
stage of Gen IV development, using oxide dispersion-strengthened steels.
They have much higher thermal conductivity than water, but lower than
sodium. Westinghouse is developing a lead-cooled fast reactor concept.
While lead has limited activation from neutrons, a problem with Pb-Bi is
that it yields toxic polonium (Po-210) activation product, an
alpha-emitter with a half-life of 138 days. Pb-Bi melts at a relatively
low 125°C (hence eutectic) and boils at 1670°C, Pb melts at 327°C and
boils at 1737°C but is very much more abundant and cheaper to produce
than bismuth, hence is envisaged for large-scale use in the future,
though freezing must be prevented. The development of nuclear power
based on Pb-Bi cooled fast neutron reactors is likely to be limited to a
total of 50-100 GWe, basically for small reactors in remote places. In
1998 Russia declassified a lot of research information derived from its
experience with submarine reactors, and US interest in using Pb
generally or Pb-Bi for small reactors has increased subsequently. The
Gen4 Module (Hyperion) reactor will use lead-bismuth eutectic which is
45% Pb, 55% Bi. A secondary circuit generating steam is likely.
SALT: Fluoride salts boil at around 1400°C at
atmospheric pressure, so allow several options for use of the heat,
including using helium in a secondary Brayton cycle circuit with thermal
efficiencies of 48% at 750°C to 59% at 1000°C, or manufacture of
hydrogen. Fluoride salts have a very high boiling temperature, very low
vapour pressure even at red heat, very high volumetric heat capacity
(carry more heat than the same volume of water), good heat transfer
properties, low neutron absorbtion, good neutron moderation capability,
are not damaged by radiation, are chemically very stable so absorb all
fission products well and do not react violently with air or water, are
compatible with graphite, and some are also inert to some common
structural metals. Some gamma-active F-20 is formed by neutron capture,
but has very short half-life (11 seconds).
Lithium-beryllium fluoride Li2BeF4 (FLiBe) salt
is a eutectic version of LiF (2LiF + BeF2) which solidifies at 459°C
and boils at 1430°C. It is favoured in MSR and AHTR/FHR primary cooling
and when uncontaminated has a low corrosion effect. LiF without the
toxic beryllium solidifies at about 500°C and boils at about 1200°C.
FLiNaK (LiF-NaF-KF) is also eutectic and solidifies at 454°C and boils
at 1570°C. It has a higher neutron cross-section than FLiBe or LiF but
can be used intermediate cooling loops.
Chloride salts have advantages in fast-spectrum
molten salt reactors, having higher solubility for actinides than
fluorides. While NaCl has good nuclear, chemical and physical
properties its high melting point means it needs to be blended with MgCl2 or CaCl2,
the former being preferred in eutectic, and allowing the addition of
actinide trichlorides. The major isotope of chlorine, Cl-35 gives rise
to Cl-36 as an activation product – a long-lived energetic beta source,
so Cl-37 is much preferable in a reactor.
All low-pressure liquid coolants allow all their
heat to be delivered at high temperatures, since the temperature drop in
heat exchangers is less than with gas coolants. Also, with a good
margin between operating and boiling temperatures, passive cooling for
decay heat is readily achieved. Since heat exchangers do leak to some
small extent, having incompatible primary and secondary coolants can be a
problem. The less pressure difference across the heat exchanger, the
less is the problem.
The removal of passive decay heat is a vital feature
of primary cooling systems, beyond heat transfer to do work. When the
fission process stops, fission product decay continues and a substantial
amount of heat is added to the core. At the moment of shutdown, this is
about 6.5% of the full power level, but after an hour it drops to about
1.5% as the short-lived fission products decay. After a day, the decay
heat falls to 0.4%, and after a week it will be only 0.2%. This heat
could melt the core of a light water reactor unless it is reliably
dissipated, as shown in 2011 at Fukushima, where about 1.5% of the heat
was being generated when the tsunami disabled the cooling. In passive
systems, some kind of convection flow is relied upon.
The top AHTR/FHR line is potential, lower one practical today. See also paper on Cooling Power Plants.
There is some radioactivity in the cooling water flowing through the
core of a water-cooled reactor, due mainly to the activation product
nitrogen-16, formed by neutron capture from oxygen. N-16 has a half-life
on only 7 seconds but produces high-energy gamma radiation during
decay. It is the reason that access to a BWR turbine hall is restricted
during actual operation.
Nuclear reactors for process heat
Producing steam to drive a turbine and generator is relatively easy,
and a light water reactor running at 350°C does this readily. As the
above section and Figure show, other types of reactor are required for
higher temperatures. A 2010 US Department of Energy document quotes
500°C for a liquid metal cooled reactor (FNR), 860°C for a molten salt
reactor (MSR), and 950°C for a high temperature gas-cooled reactor
(HTR). Lower-temperature reactors can be used with supplemental gas
heating to reach higher temperatures, though employing an LWR would not
be practical or economic. The DOE said that high reactor outlet
temperatures in the range 750 to 950°C were required to satisfy all end
user requirements evaluated to date for the Next Generation Nuclear
Plant.
Primitive reactors
The world's oldest known nuclear reactors operated at what is now
Oklo in Gabon, West Africa. About 2 billion years ago, at least 17
natural nuclear reactors achieved criticality in a rich deposit of
uranium ore. Each operated intermittently at about 20 kW thermal, the
reaction ceasing whenever the water turned to steam so that it ceased to
function as moderator. At that time the concentration of U-235 in all
natural uranium was about three percent instead of 0.7 percent as at
present. (U-235 decays much faster than U-238, whose half-life is about
the same as the age of the Earth.) These natural chain reactions,
started spontaneously with the presence of water acting as a moderator,
continued overall for about 2 million years before finally dying
away. It appears that each reactor operated in pulses of about 30
minutes – interrupted when the water turned to steam, thereby switching
it off for a few hours until it cooled. It is estimated that about 130
TWh of heat was produced. (The reactors were discovered when assays of
mined uranium showed only 0.717% U-235 instead of 0.720% as everywhere
else on the planet. Further investigation identified significant
concentrations of fission products from both uranium and plutonium.)
During this long reaction period about 5.4 tonnes of fission products
as well as up to two tonnes of plutonium together with other
transuranic elements were generated in the orebody. The initial
radioactive products have long since decayed into stable elements but
close study of the amount and location of these has shown that there was
little movement of radioactive wastes during and after the nuclear
reactions. Plutonium and the other transuranics remained immobile.
Sources: Wilson, P.D., 1996, The Nuclear Fuel Cycle, OUP. Scientific American 2005 article on Oklo Technical and Economic Aspects of Load Following with Nuclear Power Plants, OECD Nuclear Energy Agency (June 2011)
