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MEHB513 Course Project
1.0 Introduction
Many efforts and researches have been done in order to find the alternative to the electricity
generation all over the world, including our country with the increase of price in energy fuel.
There are many advantages and disadvantages of the nuclear power plant. Nuclear power plant
does not depend on the fossil fuel. Based on the statistic from World Nuclear Association, the
fuel cost to electricity ratio is lower than the fossil fuel power plant. The nuclear power plant also
does not emit CO2 as much as fossil fuel power plant does. A proper functioning nuclear power
plant actually releases fewer radio activities into the atmosphere than a coal-fired power plant.
Hence, the pollution and greenhouse effect problem can be reduced or minimized. Nuclear
power is one of the safest methods of producing energy. Based on Ralph Kinney Bennett from
University of Texas, an average of 50,000 Americans die from respiratory diseases due to the
burning of coal, and 300 are killed in mining and transportation accidents. In contrast, no
Americans have died or been seriously injured because of a reactor accident or radiation
exposure from American nuclear power plants. There are a number of safety mechanisms that
make the chances of reactor accidents to a very low level.
One of the major disadvantages of nuclear power is the risk. The major accident in nuclear
power plant can be very catastrophic and global disaster. For an example, the incident happened
in Chernobyl, which can also occur in the future. Moreover, the safety issue is always should be
emphasized because the infrastructure and building failure due to the inadequate quality of the
building in Malaysia. Furthermore, the nuclear power plant also is not a renewable and
sustainable energy. The capital cost for a nuclear power plant is very expensive and need very
high skilled workers. The waste of uranium will last its radioactivity to 10000 years before it
become stable and safe. The waste management cost can be pricey and hazard to the
environment. So, the advantages of nuclear power are what Malaysia need right now. That is
why we have to use nuclear power as long as the regulations, standards and precautions are
strictly obeyed and met.
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2.0 Objective
The nuclear power industry has undergone tremendous development and evolution for more than
half a century since the world’s first nuclear reactor major breakthrough in December 02, 1942.
After surpassing moratorium of nuclear power plant construction caused by catastrophic
accidents at Three-Mile Island (1979) and Chernobyl (1986), today, nuclear energy is back on
the policy agendas of many states, both developed and developing nations, signaling nuclear
revival or nuclear renaissance. Selection of suitable nuclear power technology has been very
crucial. We have suggested the preliminary technology assessment for the first nuclear power
reactor technology for Malaysia. Methodology employed is qualitative analysis of suitable
reactor technology from given vendors list and Preliminary Feasibility Study for Nuclear Power
Program in Peninsular Malaysia and other published presentations and/or papers by multiple
experts. The results of our research suggested that the pressurized water reactor (PWR) is the
prevailing technology in terms of numbers and plant performances, and while the
commercialization of Gen IV reactors is remote (e.g. not until 2030), Generation III/III+ NPP
models are commercially available on the market today. Moreover, five major steps involved in
reactor technology selection were introduced with a focus on introducing important aspects of
selection criteria. Recommendations for reactor technology option were also provided for both
strategic and technical recommendations. The objective of this coursework shall postulate or
rather implore what could be the best way for Malaysian and also other aspiring new entrant
nations to select systematically their first civilian nuclear power reactor.
3.0 Current 1979 National Energy Policy
Supply Objective: To ensure adequate, secure and cost-effective supply of energy.
Utilization Objective: To promote efficient utilization of energy & discourage wasteful
and non-productive patterns of energy consumption.
Environmental Objective: To ensure factors pertaining to environmental protection are not
neglected in production & utilization of energy.
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4.0 Advanced Nuclear Power Reactors
The first 3rd generation advanced reactors have been operating in Japan since 1996. Late
3rd generation designs are now being built. Newer advanced reactors have simpler designs
which reduce capital cost. They are more fuel efficient and are inherently safer. Several
generations of reactors are commonly distinguished. Generation I reactors were developed in
1950-60s, and outside the UK none are still running today. Generation II reactors are typified by
the present US and French fleets and most in operation elsewhere. Generations III (and 3+) are
the Advanced Reactors. Generation IV designs are still on the drawing board and will not be
operational before 2020 at the earliest. About 85% of the world's nuclear electricity is generated
by reactors derived from designs originally developed for naval use. These and other second-
generation nuclear power units have been found to be safe and reliable, but they are being
superseded by better designs. Reactor suppliers in North America, Japan, Europe, Russia and
elsewhere have a dozen new nuclear reactor designs at advanced stages of planning, while others
are at a research and development stage. Fourth-generation reactors are at concept stage. Third-
generation reactors have: a standardized design for each type to expedite licensing, reduce capital
cost and reduce construction time, a simpler and more rugged design, making them easier to
operate and less vulnerable to operational upsets, higher availability and longer operating life -
typically 60 years, further reduced possibility of core melt accidents,* resistance to serious
damage that would allow radiological release from an aircraft impact, higher burn-up to reduce
fuel use and the amount of waste, burnable absorbers ("poisons") to extend fuel life. The US
NRC requirement for calculated core damage frequency is 1x10-4, most current US plants have
about 5x10-5 and Generation III plants are about ten times better than this. The IAEA safety
target for future plants is 1x10-5. The large release frequency (for radioactivity) is generally
about ten times less than CDF. The greatest departure from second-generation designs is that
many incorporate passive or inherent safety features which require no active controls or
operational intervention to avoid accidents in the event of malfunction, and may rely on gravity,
natural convection or resistance to high temperatures.
Another departure is that some PWR types will be designed for load-following. While most
French reactors today are operated in that mode to some extent, the EPR design has better
capabilities. It will be able to maintain its output at 25% and then ramp up to full output at a rate
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of 2.5% of rated power per minute up to 60% output and at 5% of rated output per minute up to
full rated power. This means that potentially the unit can change its output from 25% to 100% in
less than 30 minutes, though this may be at some expense of wear and tear. Many are larger than
predecessors. Increasingly they involve international collaboration. However, certification of
designs is on a national basis, and is safety-based. In Europe there are moves towards
harmonised requirements for licensing. In Europe, reactors may also be certified according to
compliance with European Utilities Requirements (EUR) of 12 generating companies, which
have stringent safety criteria. The EUR are basically a utilities' wish list of some 5000 items
needed for new nuclear plants. Plants certified as complying with EUR include Westinghouse
AP1000, Gidropress' AES-92, Areva's EPR, GE's ABWR, Areva's Kerena, and Westinghouse
BWR 90. European regulators are increasingly requiring large new reactors to have some kind of
core catcher or similar device, so that in a full core-melt accident there is enhanced provision for
cooling the bottom of the reactor pressure vessel or simply catching any material that might melt
through it. The EPR and VVER-1200 have core-catchers under the pressure vessel, the AP1000
and APWR have provision for enhanced water cooling. In the USA a number of reactor types
have received Design Certification (see below) and others are in process such as ESBWR from
GE-Hitachi, US EPR from Areva and US-APWR from Mitsubishi. The ESBWR is on track to
receive certification about September 2011, and the US EPR in mid 2012. Early in 2008 the
NRC said that beyond these three, six pre-application reviews could possibly get underway by
about 2010. These included: ACR from Atomic Energy of Canada Ltd (AECL), IRIS from
Westinghouse, PBMR from Eskom and 4S from Toshiba as well as General Atomics' GT-MHR
apparently. However, for various reasons these seem to be inactive. Longer term, the NRC
expected to focus on the Next-Generation Nuclear Plant (NGNP) for the USA essentially the
Very High Temperature Reactor (VHTR) among the Generation IV designs.
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5.0 Screening Criteria of Reactor Technologies.
5.1 The maximum power rating of the first NPP to be constructed is 2000 MWe
For a high power generation, Pressurized Water Reactor (PWR) is the best choice. There are
many advantages of PWR as compared to Boiler Water Reactor (BWR) that suitable for design
with four first-stage coolant circulation loops per reactor. First of all, PWR reactors are very
stable due to their tendency to produce less power as temperatures increase. This makes the
reactor easier to operate from a stability standpoint. Next, PWR turbine cycle loop is separate
from the primary loop, so the water in the secondary loop is not contaminated by radioactive
materials. Next, PWRs can passively scram the reactor in the event that offsite power is lost to
immediately stop the primary nuclear reaction. The control rods are held by electromagnets and
fall by gravity when current is lost; full insertion safely shuts down the primary nuclear reaction.
Moreover, PWR technology is favored by nations seeking to develop a nuclear navy, the
compact reactors fit well in nuclear submarines and other nuclear ships as compared to BWR.
For selection of type of generation type design, it is preferable to use the third generation. This is
because third generation type will ensure fast and sure termination of the nuclear reaction in the
reactor core thanks to two individual completely independent reactivity control systems.
Redundancy for all safety functions provided by the use of both active and passive safety
systems (including a Passive Residual Heat Removal System and Passive Filtering System),
which require neither operator intervention nor electric power supply. Besides that, it has special
enclosure to contain potential accidents. This structure is composed of a primary containment of
pre-stressed reinforced concrete and a leak tight metal liner, secondary reinforced concrete
containment, and cast concrete external structure designed to withstand a large range of internal
and external events. However, there is some conflict of constructing this NPP in Malaysia. It
brings the negative effect on the tourist industry and the agriculture. Other than that, the problem
is safety concerns over the use of nuclear technology, location in a seismic active zone, the
expense of the project, risk of terrorist attack, problems with the transportation, procession and
preservation of the nuclear waste and also complete unnecessity of further nuclear power in the
first place, as better options are available.
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5.2 The purpose of NPP to be introduced in the country is for the electricity generation and
the economy-of-scale calls for large capacity NPP’s with power rating greater than 500
MWe.
Importance of electricity are important to our lives. All it takes is a power failure to remind us
how much we depend on it. Life would be very different without electricity as people and the
industry depend on it.. It is the responsibility of electric utility companies to make sure
electricity is there when we need it. They must consider reliability, capacity, base load, peak
demand, and power pools. Reliability is the capability of a utility company to provide electricity
to its customers 100 percent of the time. A reliable electric service is without blackouts or
brownouts. To ensure uninterrupted service, laws require most utility companies to have 15 to 20
percent more capacity than they need to meet peak demand. This means a utility company whose
peak load is 12,000 megawatts (MW) must have 14,000 MW of installed electrical capacity.
There will be enough electricity to meet demand even if equipment were to break down on a hot
summer afternoon. Capacity is the total quantity of electricity a utility company has on-line and
ready to deliver when people need it.. A utility company that has seven 1,000 MW plants, eight
500 MW plants, and 30 100 MW plants has a total capacity of 14,000 MW. Base load power is
the electricity generated by utility companies around-the-clock, using the most inexpensive
energy sources usually coal, nuclear, and hydropower. When many people want electricity at the
same time, there is a peak demand. Power companies must be ready for peak demands so there is
enough power for everyone. These peak load generators run on natural gas, diesel, or
hydropower and can be put into operation in minutes. The more this equipment is used, the
higher our utility bills. By managing the use of electricity during peak hours, we can help keep
costs down. The second question on generating electricity is how much does electricity cost? The
answer depends on the cost to generate the power (60 percent), the cost of transmission (8
percent), and local distribution (32 percent). The average cost of electricity is twelve cents per
kWh for residential customers and a little under seven cents for industrial customers. A major
key to cost is the fuel used to generate the power. Electricity produced from natural gas, for
example, costs more than electricity produced from uranium, coal, or hydropower. Another
consideration is how much it costs to build a power plant. A plant may be very expensive to
construct, but the cost of the fuel can make it competitive to other plants, or vice versa. Nuclear
power plants, for example, are very expensive to build, but their fuel uranium is very cheap. A
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coal-fired plant, on the other hand, is much less expensive to build than nuclear plants, but their
fuel coal is more expensive. When calculating costs, a plant’s efficiency must also be considered.
In theory, a 100 percent energy efficient machine would change all the energy put into the
machine into useful work, not wasting a single unit of energy. But converting a primary energy
source into electricity involves a loss of usable energy, usually in the form of heat. In general, it
takes three units of fuel to produce one unit of electricity from a thermal power plant. Today’s
thermal power plants are over eight times more efficient with efficiency ratings around 35
percent. Still, this means 65 percent of the initial heat energy used to make electricity is lost. A
modern coal plant burns about 5,000 tons of coal each day, and about two-thirds of this is lost
when the chemical energy in coal is converted into thermal energy, then into electrical energy. A
hydropower plant is about 90 percent efficient at converting the kinetic energy of moving water
into electricity. But that’s not all. About two percent of the electricity generated at a power plant
must be used to run equipment. And then, even after the electricity is sent over electrical lines,
another seven percent of the electrical energy is lost in transmission. Of course, consumers pay
for all the electricity generated, lost or not. The cost of electricity is affected by what time of day
it is used. During a hot summer afternoon from noon to 6 p.m., there is a peak of usage when air-
conditioners are working harder to keep buildings cool. Electric companies charge their
industrial and commercial customers more for electricity during these peak load periods because
they must turn to more expensive ways to generate power. Nuclear plants generate electricity
much as fossil fuel plants do, except that the furnace is a reactor and the fuel is uranium. In a
nuclear plant, a reactor splits uranium atoms into smaller elements, producing heat in the
process. The heat is used to superheat water into high-pressure steam, which drives a turbine
generator. Like fossil fuel plants, nuclear power plants are thermal plants because they use heat
to generate electricity. Malaysia should consider to build NPP for generating electricity as the
future estimation indicates that about 223 more gigawatts by 2035, which is predicted by experts.
Among all of the electricity generating sources, nuclear can said to be the relevant way to
generate electricity in Malaysia because it provides diversification and security of supply and
nuclear is an economically viable base load option and also provides power without carbon
emission.
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5.3 The national natural resource utilization strategy of the country does not preclude the
use of enriched uranium fuels available in the world market.
Nuclear power is generated from a metal called uranium which mined in places such as
Australia, Argentina, Canada, Brazil, China, France and many other countries. The uranium is
used to create controlled nuclear reactions called nuclear fission. In a standard fossil fuel
electricity station the same thing happens except the burning of the fossil fuel heats the water to
create steam. Nuclear power plants can give more energy using less fuel and thus a more
environmental friendly source of power. However, both the pros and cons of nuclear power
plants are to be considered as the incident in Japan at Fukushima cannot be kept aside. The
production of nuclear power produced a lot of waste material which hazardous and difficult to
dispose of safely. Most radioactive waste produced by a nuclear power plant is keep on site or is
buried underground in huge concrete pits. This waste has to be carefully looked after for many
thousands of years. While the chance of a nuclear accident happening are relatively low the risks
associated are extremely high. A small radiation leaks can give the people and enviroment a
devasting memory.For an example, a cataclysmic accident can cause widespread radiation
poisoning across many countries as the Chernobyl Nuclear disaster proved. Due to the wide
spread damage a reactor breach could cause, nuclear power plants could become favoured
terrorist targets because weapon can be created by using the plotinum that can be found in
abundance from nuclear fission. Nuclear power is not a renewable source of energy. Uranium is
a type of metal that can be found in mining area. It is a scarce metal and the supply of uranium
will one day run out making all the nuclear power plants obsolete. Nuclear power is not a long
term solution to finding a renewable, environmentally friendly energy source. To make a nuclear
plant and making sure it is effective takes a long time.. In addition to the time it take to
effectively plan and build a nuclear power plant, a lot of money has to be spent to make sure that
the plant is safe. Unsafe plants mean that there are more likely to be accidents. So, it is a wise
decision that the national natural resource utilization strategy of the country does not include the
use of uranium fuels that is available in the market.
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5.4 The reactor technology is proven and standardized if the technology satisfies one of the
followings;
• The same design is in operation or under construction
• The same design is design certified by the regulatory authority of the country of
origin
5.4.1 Man-Machine Interface System
Man-Machine Interface System (MMIS) design shall be one of key factors for evaluation
ofreactor technology selection since NPP is operated by plant personnel. All aspects of
plantdesign which require interfacing with plant personnel shall incorporate human
factorsconsiderations. Human factors driven design features shall be applied consistently plant-
wide.Amongst top-level requirements for MMIS include:
- Use of modern digital technology, including multiplexing and fiber optics, for
monitoring, control, and protection functions.
- Segmentation and separation on safety and protection systems.
- Use of compact, redundant, operator work stations with multiple display and control
devices that provide organized, hierarchical access to alarms, displays, and controls.
- Incorporate modern, computer-driven displays to provide enhanced trending information,
validated data, and alarm prioritization and supervision, as well as diagrammatic normal,
abnormal, and emergency operating procedures with embedded dynamic indication and
alarm information.
- Include large, upright, spatially dedicated panels which provide an integrated plant
mimic, indicating equipment status, plant parameters, and high level alarms.
- Lighting levels, HVAC, sound levels, colors, etc., shall provide a comfortable,
professional atmosphere that enhances operator effectiveness and alertness.
- Local and stand-alone control systems shall be designed in the same rigorous way as the
main control stations and will use consistent labeling, nomenclature, etc. Particular
attention is to be paid to visibility, color coding, use of mimics, access, lighting, and
communication.
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- An integrated, plant wide communications system shall be provided for construction and
operations.
5.4.2 Operability, Maintainability and Testing
Important aspects for operability, maintainability, and testing of NPP design are as follows:
- Ease of operation shall be achieved through the use of modem digital technology for
monitoring, control, and protection functions, a forgiving plant response to upset
conditions, design margins, and consideration on the operating environment.
- Experience feedback of O&M problems which exist in current plants.
- Minimize the number of different types of equipment by standardization except for those
limited applications where diversification is adopted to protect against common mode
failure (CMF).
- Design to facilitate replacement of major components such as steam generators, within
design availability limits.
- Equipment design to have minimal, simple maintenance needs, and be designed to
facilitate needed maintenance.
- Consideration of the maintenance access, pull and laydown space, and heavy lifts.
- Environmental design to provide satisfactory working conditions, including temperature,
dose, ventilation, and illumination.
- Design to facilitate the use of robots addressing arrangements to accommodate
movement, access ports, communication, and robot storage and decontamination.
- The surveillance tests shall be designed to measure the systems design basis performance
parameters, preferably with the plant at power in order to avoid adding tasks to
theplanned outage time. Mechanical and electrical systems shall be designed to avoid
plant trips, and plant equipment and to facilitate and simplify surveillance testing.
- The protection system and control systems for the engineered safety systems shall be
designed so that the plant can be safely operated indefinitely at full power with one
protection channel in test or bypassed one subsequent single failure will not cause a plant
trip.
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5.5 The country prefers the advanced reactor technologies with the safety and performance
goals are set to be equal or equivalent to those of the US Electric Power Research Institute
(EPRI) ALWR Utility Requirement Documents (URD).
The safety systems include passive safety injection, passive residual heat removal, and passive
containment cooling. All these passive systems meet the NRC single-failure criteria and
addresses Three Mile Island lessons learned, unresolved safety issues, and generic safety issues.
These passive systems have been proven through extensive testing and computer code analysis at
two different power levels (AP600and AP1000).Passive systems and the use of experience-based
components do more than increase safety, enhance public acceptance of nuclear power, and ease
licensing—they also simplify overall plant systems, equipment, and operation and maintenance.
The simplification of plant systems, combined with large plant operating margins, greatly
reduces the actions required by the operator in the unlikely event of an accident. Passive systems
use only natural forces, such as gravity, natural circulation, and compressed gas-simple physical
principles we rely on every day. There are no pumps, fans, diesels, chillers, or other rotating
machinery required for the safety systems. This eliminates the need for safety-related AC power
sources. A few simple valves align the passive safety systems when they are automatically
actuated. In most cases, these valves are “fail safe.” They require power to stay in their normal,
closed position. Loss of power causes them to open into their safety alignment. In all cases, their
movement is made using stored energy from springs, compressed gas or batteries. Simple
changes in the safety-related systems from AP600 to AP1000 allow accommodation of the
higher plant power without sacrificing design and safety margins. Since there are no safety
related pumps, increased flow was achieved by increasing pipe size. Additional water volumes
were achieved by increasing tank sizes. These increases were made while keeping the plant
footprint unchanged. This ensures that the designs of other systems are not affected by layout
changes. Enforcing a rigorous “no unnecessary change policy” makes that portion of the detail
design also complete.
5.5.1 Passive safety system
Passive systems provide plant safety and protect capital investment. They establish and maintain
core cooling and containment integrity indefinitely, with no operator or AC power support
requirements. The passive systems meet the single failure criteria and probabilistic risk
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assessments (PRA) used to verify reliability. The passive safety systems are significantly simpler
than typical PWR safety systems. They contain significantly fewer components, reducing
required tests, inspections, and maintenance. The passive safety systems have one-third the
number of remote valves as typical active safety systems, and they contain no pumps. Equally
important, passive safety systems do not require a radical departure in the design of the rest of
the plant, core, RCS, or containment. The passive safety systems do not require the large
network of active safety support systems needed in typical nuclear plants. These include AC
power, HVAC, cooling water, and the associated seismic buildings to house these components.
This simplification applies to the emergency diesel generators and their network of support
systems, air start, fuel storage tanks and transfer pumps, and the air intake/exhaust system. These
support systems no longer must be safety class, and they are either simplified or eliminated. For
example, the essential service water system and its associated safety cooling towers are replaced
with a non-safety-related service water cooling system. Non-safety-related support systems and
passive safety systems are integrated into the plant design. Licensing safety criteria are satisfied
with a greatly simplified plant.
5.5.2 Emergency Core Cooling System
The passive core cooling system (PXS), protects the plant against RCS leaks and ruptures of
various sizes and locations. The PXS provides core residual heat removal, safety injection, and
depressurization. Safety analyses (using NRC-approved computer codes) demonstrate the
effectiveness of the PXS in protecting the core following various RCS break events. Even for
breaks as severe as the 20.0 cm (8 in.) vessel injection lines, there is no core uncover for either
AP600or AP1000. Following a double-ended rupture of a main reactor coolant pipe, the PXS
cools the reactor with ample margin to the peak clad temperature limit.
5.5.3 Safety injection and depressurization
The PXS uses three sources of water to maintain core cooling through safety injection. These
injection sources include the core makeup tanks (CMTs), the accumulators, and the in
containment refueling water storage tank (IRWST). These injection sources are directly
connected to two nozzles on the reactor vessel so that no injection flow can be spilled in case of
larger breaks. Long-term injection water is provided by gravity from the IRWST, which is
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located in the containment just above the RCS loops. Normally, the IRWST is isolated from the
RCS by squib valves and check valves. This tank is designed for atmospheric pressure. The RCS
must be depressurized before injection can occur. The RCS is automatically controlled to reduce
pressure to about 0.83 bar (12 psig), at which point the head of water in the IRWST overcomes
the low RCS pressure and the pressure loss in the injection lines. The PXS provides
depressurization using the four stages of the automatic depressurization system (ADS) to permit
a relatively slow, controlled RCS pressure reduction. To maintain similar margins for accidents
requiring safety injection, a few lines in the PXS were made larger for AP1000. In addition, the
CMTs were enlarged to provide adequate margin without requiring redesign of adjacent piping
and structure.
5.5.4 Passive residual heat removal
The PXS includes one passive residual heat removal heat exchanger (PRHR HX). The PRHR
HX is connected through inlet ad outlet lines to RCS loop 1. The PRHR HX protects the plant
against transients that upset the normal steam generator feedwater and steam systems. It satisfies
the safety criteria for loss of feedwater, feedwater line breaks, and steam line breaks. For
AP1000, the PRHR HX horizontal tube portions were made slightly longer and a few tubes were
added to the existing AP600 PRHR HX tube sheet. PRHR piping was made larger. These
modifications resulted in a 100% capacity system without affecting surrounding piping and
layout design. The IRWST provides the heat sink for the PRHR HX. The IRWST water absorbs
decay heat for more than one hour before the water begins to boil. Once boiling starts, steam
passes to the containment. The steam condenses on the steel containment vessel and, after
collection, drains by gravity back into the IRWST. The PRHR HX and the passive containment
cooling system provide indefinite decay heat removal capability with no operator action
required. For AP1000 the normal water level in the IRWST was raised to provide adequate water
inventory without changing the structure.
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5.6 The vendors will promote the largest capacity model among their available design
portfolio for the country.
The vendors will promote the largest capacity of their model by giving extra information to the
company that going to construct the nuclear power plant. Among the vendors technique are by
promoting the advantages of their available model in energy saving and effectiveness in
generating the power. Besides that the vendors could also promote their nuclear reactor by the
way the nuclear reactor capacity in their fuel. As an example the GE Hitachi advanced boiling
water reactor (ABWR) of 1300-1500 MWe. Several ABWRs are now in operation in Japan, with
more under construction there and in Taiwan. Some of these have had Toshiba involved in the
construction, and it is now Toshiba that is promoting the design most strongly in the USA.
6.0 Lists of Reactor Technologies for Selection.
6.1 Korea –KEPCO Consortium
Reactor Type: APR-1400
Reactor Capacity: 1450 MWe
The APR1400 currently being marketed for export by KEPCO has had added to its design
significant enhancements in regard to safety as well as increased power capabilities. Based upon
the predecessor OPR1000 reactor and Korea’s experience gained over the country’s non-stop
development of nuclear reactors, the upgraded APR1400 has been designed to utilize the proven
technology of the earlier model while offering more in terms of safety, performance,
construction period, operation and of course, economics.
By adopting advanced design features based on self-reliant technologies as well as on the
technologies of the System 80+, whose design was certified by the Nuclear Regulatory
Commission, Korea developed the APR1400 to meet the Korean Utility Requirement Document
(KURD) reflecting the Advanced Light Water Reactor (ALWR) design requirements developed
by the Electric Power Research Institute (EPRI) and other nuclear power related bodies. The
lifespan of the APR1400 reactor was also increased to 60 years, 20 years longer than its
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OPR1000 predecessor, a reactor which was developed as an integral part of Korea’s NPP
standardization program begun in 1984. Ulchin Unit 3 was the first OPR1000 to go into
operation in 1998, three years after Korean NPPs had reached a level of 95 percent indigenous
technology.
6.1.1 Features of the APR-1400 include:
Advanced Design
4-train direct vessel injection safety system and fluidic device in safety injection tank
In-containment refuelling water storage tank
Digital I&C and operator-friendly man-machine interface
Improved Cost Effectiveness
Extended plant design lifetime
Reduced operation & construction cost
Minimum site boundary via plant general arrangement optimization
State-of-the-art construction technologies
Enhanced Safety
Adoption of proven & evolutionary technologies
Reduced core damage & containment failure frequency
Reinforced seismic design basis
Improved severe accident mitigation system
Convenient Operation & Maintenance
Extended operator response time
Reduced occupational exposure
Convenient facilities for improved maintenance & inspection
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The usage of Uranium fuel.
The nuclear fuel for the APR1400 was designed by Korea Nuclear Fuel. The APR1400 uses
uranium dioxide, processed from enriched uranium-235. To manufacture the nuclear fuel, the
uranium dioxide must first be processed into a powder form. It is then pressed into cylindrical
pellets of approximately 10mm in length, 8mm in width, and weighing 5.2 grams.
Approximately 365 pellets are then stacked end-to-end inside a hollow fuel rod made of a
zirconium-niobium alloy. To get an idea of the power output by the APR1400, one uranium
dioxide pellet produces about 1,600 KWh of electricity, which is the average amount of
electricity that one household uses over an 8 month period. Each fuel rod contains 365 pellets,
and one PLUS7 fuel assembly is made up of 236 fuel rods. This means that each fuel assembly
contains about 86,140 uranium dioxide pellets. Since there are 241 fuel assemblies in each
APR1400 reactor vessel, one APR1400 reactor can generate enough electricity to power roughly
13.3 million homes for one year.
The reactor is proven and standardized.
The nuclear power industry has been developing and improving reactor technology for more than
five decades and is starting to build the next generation of nuclear power reactors to fill new
orders. 80 reactors in South Korea incorporate many design features of the System 80+ advanced
pressurised water reactor (PWR), which is the basis of the Korean Next Generation Reactor
program, specifically the APR-1400 which is expected to be in operation from 2013 and is being
marketed worldwide.
US EPRI and ALWR URD.
The Advanced Power Reactor 1400 (APR1400) design is an evolutionary ALWR design and
incorporates a variety of engineering improvements and operational experience to enhance
safety, economics and reliability. Design features to address the NRC’s Severe Accident and
Safety Goal Policy Statements are also incorporated into the APR1400 design. The APR1400
design is based on the actual experience from the OPR1000 design, configuration of the reactor
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coolant system (RCS) of the APR1400 is identical to that of the OPR1000. Advanced design
features and improvements have been incorporated: a pilot operated safety relief valve (POSRV),
a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the
safety injection tank, IRWST, an external reactor vessel cooling system, and an integrated head
assembly (IHA). The evolutionary APR1400 design concept is based on Korean Utility
Requirements Document (KURD) that was established by making reference to ALWR utility
requirements documents developed by Electric Power Research Institute and organizations in
other countries. The Central Research Institute of Korea Hydro and Nuclear Power Co. Ltd.
announced that, in cooperation with Electric Power Research Institute (EPRI) of the United
States, it has completed an establishment of APR1400 Material Management System as an effort
to enhance integrity of the main reactor significantly.
The Nuclear Power Plant Material Management System is a tool that allows operators to
systemically manage deterioration of materials as well as material quality of the main reactor,
which are directly related to the safety of a nuclear power plant throughout its life cycle, from
design and construction to operation. As the system contains data such as material names,
deterioration mechanism, threat levels, design and construction standards as well as solutions for
relevant problems, it not only provides all the necessary information but also supports in coming
up with pre-emptive measures for prevention of the material deterioration throughout a nuclear
power plant's service life from its design to operation.
Whereas AP1000, APWR, US EPR, and other competing reactors have already had their own
well-established material management systems, APR1400 has yet to have its own material
management system in place until now. Therefore, the material management system of APR1400
is critical for the reactor to be recognized of its operational reliability so much as for global
reputation, while it is also expected to enhance its competitive edge in the global market. This
satisfy the criteria which the safety and performances are set equivalent to US EPRI and ALWR
URD.
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Comparison of OPR1000 and APR1400 .
Parameters OPR1000 APR1400
Thermal/Elec. Power 2825MWt / 1000MWe 4000MWt / 1450MWe
Design Life 40Yrs 60Yrs
Seismic Acceleration 0.2g 0.3g
Safety Requirements
- CDF
- Thermal Margin
- Operator Action Time
- Emergency Core Cooling
< 10-4/RY
8%
Min. 10 minutes
2-train, Cold leg Injection
< 10-5/RY
10%
Min. 30 minutes
4-train, DVI, Fluidic Device in
SIT
Performance Requirements
- Plant Availability
- Unplanned Trip
- Refuelling Cycle
87%
<1/yr
15-18 months
90%
<0.8/yr
18-24 months
MMIS Digital Digital
Others
- Reactor Vessel Wall
Cooling
- RWST
Air Cooling
Outside Containment
ERVC
Inside Containment
Design Concept of OPR1000
Design optimization utilizing proven technology
Feedback of operating experiences
Improved plant economics and safety
Standardized design
Design Improvements that has been made in APR1400
Design improvements to meet ALWR URD
High reliability and better performance
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4 train safety injection, DVI, IRWST
POSRV for stable operation
Severe accident mitigation: ERVC
Digital protection and control systems
This satisfies the criteria stating that the vendor promotes the latest capacity model among their
available design for the specific country.
6.2 Country: Japan
Vendor: Mitsubishi
Reactor Type: US-APWR (PWR)
Reactor capacity: 1700MWe
The US-APWR, developed especially for use in the United States, is an economically efficient,
reliable, safe, and well-proven 1,700 MW class plant based on the advanced pressurized water
reactor (APWR), which has been already developed in Japan. It utilizes the latest and best
technology such as high-performance steam generators and turbines complies with US regulatory
requirements such as enhanced safety power systems, reflects US customer needs such as
reduced generation costs for long cycle operation, and applies US site conditions which enable a
compact building layout based on moderate seismic conditions.
Mitsubishi's large APWR (advanced PWR) of 1538 MWe gross (4451 MWt) was developed in
collaboration with four utilities (Westinghouse was earlier involved). The first two are planned
for Tsuruga, coming on line from 2016. It is a 4-loop design with 257 fuel assemblies and
neutron reflector, is simpler, combines active and passive cooling systems in double
containment, and has over 55 GWd/t (and up to 62 GWd/t) fuel burn-up. It is the basis for the
next generation of Japanese PWRs. The planned APWR+ is 1750 MWe and has full core MOX
capability.
The US-APWR will be 1700 MWe gross, about 1620 MWe net, due to longer (4.3m) fuel
assemblies, higher thermal efficiency (37%) and has 24 month refuelling cycle. Its emergency
core cooling system (ECCS) has four independent trains, and its outer walls and roof are 1.8 m
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thick. US design certification application was in January 2008 with approval expected in mid
2013 and certification in 2014. In March 2008 MHI submitted the same design for EUR
certification, as EU-APWR, and it will join with Iberdrola Engineering & Construction in
bidding for sales of this in Europe. Iberdrola would be responsible for building the plants.
The main specifications of the US-APWR are:
Electric Power 1,700 MWe
Core Thermal Power 4,451 MWt
Reactor Fuel Assemblies 257
Reactor Fuel Advanced 17x17, 14 ft.
Active Core Length 4.2 meters
Coolant System Loops 4
Coolant Flow 2.75x104 m3/h/loop
Coolant Pressure 15.5 MPa
Steam Generator Type 90TT-1
Number of Steam Generators 4
Reactor Coolant Pump Type 100A
Number of Reactor Coolant
Pumps
4
Reactor Coolant Pump Motor
Output
6,000 kW
6.2.1 Features of the US-APWR include:
Enhanced Safety:
A four-train safety system for enhanced redundancy.
An advanced accumulator.
An in-containment refuelling water storage pit.
Enhanced Reliability:
A steam generator with high corrosion resistance.
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A neutron reflector with improved internals.
A 90% reduction in plant shutdowns compared to other four-loop PRWs.
Attractive Economy
A large core with a thermal efficiency of 39%.
Building volume per MWe that is four-fifths that of other four-loop PWRs.
More Environment Friendly
A 28% reduction in spent fuel assemblies per MWh compared to other four-loop PWRs.
Reduced occupational radiation exposure.
Capacity to use mixed oxide (MOX) fuels made from reprocessed nuclear fuel waste.
Natural resources utilization
The US-APWR fuel system design inherits the reliable features which have been confirmed by
the extensive irradiation experience of Mitsubishi fuel. The latest design features are adapted to
the US-APWR fuel design, such as higher density pellet of 97%TD, gadolinia-uranium dioxide
fuel with gadolinia of up to 10wt%, ZIRLO cladding and the counter measure design to debris
fretting, grid fretting and incomplete control rod insertion.
US-APWR Technology
Mitsubishi's US-Advanced Pressurized Water Reactor (US-APWR) design is more efficient than
any previous power plant. This design has been slightly modified to satisfy U.S. and
international utility requirements, and is currently being reviewed by the Nuclear Regulatory
Commission. We believe it is important to continually build on previous progress to make
nuclear energy safer, more efficient and more environmentally friendly.
The advanced technology with the safety and performance set based on ALWR URD
The US-APWR is an advanced light water reactor plant designed by Mitsubishi Heavy
Industries, Ltd. The US-APWR reactor is a 4-loop pressurized water reactor (PWR) and has a net
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electrical power rating of approximately 1600 MWe, depending on site conditions. The rated
core thermal power level of the US-APWR is 4451 MWt.
The Reactor Building, the Power Source Buildings, the power source fuel storage vaults, and the
essential service water pipe tunnel are designed and constructed as safety-related structures, to
the requirements of seismic Category I. These safety-related structures are designed for the
effects of all applicable loads and their combinations, including the postulated seismic response
loads. These structures are designed to withstand the effects of such natural phenomena such as
hurricanes, floods, tornados, tsunamis, and earthquakes without loss of capability to perform
their safety functions. They are also designed to withstand the effects of postulated internal
events such as fires and flooding without loss of capability to perform their safety functions.
Vendor’s reactors capacity
1. U.S Current 4 Loop -------> 1,180MWe
2. APWR -------> 1,538MWe
3. US-APWR ---------> 1700MWe
Additional features of US-APWR compared to APWR as follows:-
1700MWe class output is achieved from a 10% higher efficiency than APWR.
• Same core thermal output with APWR
• High-performance, large capacity steam generator
• High-performance turbine
Low power density core using 14ft. fuel assemblies with the same reactor vessel as
APWR to enhance fuel economy for 24 months operation.
Enhanced reliability and maintainability of reactor vessel by top mounted ICIS.
Enhanced safety by 4 train safety electrical systems.
Enhanced on line maintenance capability.
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6.3 Country: USA-JAPAN
Vendor: GE-HITACHI
Reactor: ESBWR
Reactor Capacity: 1550MWe
GE Hitachi Nuclear Energy's ESBWR (Economic Simplified Boiling-Water Reactor) is an
improved design that utilizes passive safety features and natural circulation principles and is
essentially an evolution from a predecessor design, the SBWR at 670 MWe. GE-H says it is safer
and more efficient than earlier models, with 25% fewer pumps, valves and motors, and can
maintain cooling for six days after shutdown with no AC or battery power. The emergency core
cooling system has eliminated the need for pumps, using passive and stored energy.
The ESBWR (4500 MWt) will produce approximately 1600 MWe gross, and 1535 MWe net,
depending on site conditions, and has a design life of 60 years. It was more fully known as the
Economic & Simplified BWR (ESBWR) and leverages proven technologies from the ABWR.
The ESBWR is in advanced stages of licensing review with the US NRC for GE Hitachi and is
on schedule for full design certification in 2012. Design approval was in March 2011. It was
submitted for UK Generic Design Assessment in 2007, but a year later GE-H requested that this
be suspended.
GEH is selling this alongside the ABWR, which it characterises as more expensive to build and
operate, but proven. ESBWR is more innovative, with lower building and operating costs and a
60-year life.
6.3.1 Features & Benefits
Simplified design builds on proven technology, increasing fuel efficiency yet decreasing
overall operation and maintenance costs
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Sophisticated control systems – fully digital, providing reliable and accurate plant
monitoring, control, and diagnostics
Enhanced safety compliance with three layers of protection during an accident and the
highest regulatory ratings
Expedited, economical construction schedule due to pre-licensed design and standardized
modules
Experienced global supply chain team with referenced construction schedule of 36
months
Environmentally friendly with nearly zero greenhouse gas emissions equal to taking 1.5
million cars off the road
Currently in the U.S. Design Certification process
6.3.2 Design
The ESBWR is a simplified design that builds on the inherent advantages of BWR
designs and the proven operation at BWRs Dodewaard and Humboldt Bay, where natural
circulation designs have been implemented. The passive safety design features of the
ESBWR rely on natural forces like gravity, evaporation, and condensation rather than
'active' systems that rely on pumps and valves to ensure safety in the event of a
malfunction.
Simplified passive natural circulation design eliminates 11 systems and 25 percent of
pumps, valves, and motors from previous designs decreasing maintenance and increasing
worker safety.
BWR features, including isolation condensers, natural circulation, and debris-resistant
fuel, are operationally proven and provide multiple levels of protection for the outside
environment.
Very reliable passive Emergency Core Cooling System provides a large margin for loss
of cooling accidents.
Full engineering before on-site work and standardized modules result in expedited
construction schedule.
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The advanced construction techniques and management philosophies deployed in
construction of 8 Generation III ABWR plants are currently being integrated into both the
design and construction plan of the ESBWR.
Incorporating current construction experience and modularization into the design is the
foundation for a reference construction schedule of 36 months for the ESBWR.
6.3.3 Specifications
Output Power: 1520 MW net
Emissions: Nearly zero greenhouse gas emissions
Lifetime: 60 years
Core Damage Frequency: 2 E-8, the safest in the industry by a factor of ten
Availability and Capacity Factor: 95% with a capacity factor greater than the current
BWR fleet's average of 92%
Cycle Length: 24-month cycles GEH BWRs represent the top six longest running light
water reactors. Half of the top twenty longest running reactors are GEH BWRs owned by
5 independent utilities. ESBWR builds on this proven operational superiority of the
BWR.
Natural Resource Utilization
On September 25, 2012—GE Hitachi Nuclear Energy (GEH) Global Laser Enrichment (GLE)
announced receipt of its license from the U.S. Nuclear Regulatory Commission (NRC) to build a
groundbreaking laser enrichment facility on the 1,600-acre site of the company’s global
headquarters in Wilmington, N.C.
While a commercialization decision must still be made by the company, the license enables GLE
to build a first-of-its-kind uranium enrichment facility using lasers conceived of by Australian
technology company SILEX and developed by GLE experts. The company has worked with the
NRC, the U.S. departments of State and Energy and independent non-proliferation experts for
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several years to ensure the security of this technology and has met and in many cases exceeded
all regulations pertaining to safeguarding this technology.
Today, a majority of enriched uranium made to produce nuclear fuel in the United States comes
from foreign or government-supplemented sources. The GLE license, applied for in June 2009,
will allow the laser enrichment plant to produce up to 6 million single work units (SWU) per
year in the United States.
Advanced reactor technologies with safety and performance goal
GE Hitachi Nuclear Energy (GEH) brought the next-generation reactor model, the Economic
Simplified Boiling Water Reactor (ESBWR), has passed a crucial safety review performed by an
advisory committee for the U.S. Nuclear Regulatory Commission (NRC). Completion of this
review clears a key hurdle in the company’s bid for design certification of the ESBWR, which
begins the federal rulemaking process. This sets the stage for final NRC certification by the fall
of 2011. The 1,520-megawatt (MW) ESBWR offers what GEH believes is the world’s most
advanced passive safety features, simplified construction and operation and the lowest core
damage frequency on the market today. In addition, the ESBWR’s innovative digital
instrumentation and control design and development process are rigorously compliant to nuclear
regulations and globally recognized standards.
Probabilistic risk assessment.
The ESBWR design certificate included a PRA (probabilistic risk assessment) in according with
regulatory requirements. The ESBWR PRA is level 3 PRA covers full power operation and shut
down conditions. The scope of initiating events includes internal events and assessment of
internal plant fires and floods. The only quantified external events are high winds and tornadoes.
A seismic margin analysis was performed, but risk from seismic events and other possible
external events were not quantified. Although many of the analysis elements are consistent with
AS ME-RA-Sb-2005 capability 2 standard, those attributes were not consistently achieved at this
stage of the PRA development. For example some aspects of human performance, models for
equipment lasting and maintenance and details of fire and flood damage cannot be anal sized I
the absence of physical plant, procedures and operating staff. In these cases surrogate analysis
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were performed and assumptions were applied to encompass potential plant configuration,
operations and maintenance program and organization. In addition any analysis requiring site
specific characteristics were treated in a generic manner. Over view found that this PRA was
acceptable for design certification purposes. The estimated frequencies of core damage and large
releases provide confidence that ESBWR design achieves NRC staff expectations for advanced
plants. The PRA was an integral part of the ESBWR design process, and risk unsightly
influenced a number of design changes though the review. The integrated risk perfectibility was
an important contribution to achieving the estimated low -risk. The ESBWR design is robust and
there is reasonable assurance that it can be built and operated without undue risk to the health
and safety of the public
Safety enhancement features in ESBWR.
The ESBWR is direct cycle, natural circulation BWR and has passive safety features to cope up
with range of design basis accidents (DBAs). Within the containment structure are the isolation
condensers (IC) to be elevated gravity driven cooling systems (GDCS) water pools, a passive
containment cooling system (PCCS) and an elevated suppression pool. These systems can
remove decay heat under all conditions. The ESBWR standard design includes a reactor building
that surrounds the containment, as well as building dedicated exclusively or primary to housing
related systems and equipment.
The limiting ESBWR DBA is a main stream line break (MSLB). In this DBA water and steam
are initially discharged from break into dry well. As the dry well pressure increases the
horizontal vents between dry well and wet well clear. Subsequently, a steam water mixture from
break flows through the vents into wet well suppression pool, where steam is condensed and
water is cooled to the pool temperature. As primary system pressure fall to the dry well pressure,
water makes up to the reactor vessel is provided by actuation of GDCS. Example, GDCS squib
valves open and water flows by gravity head into the vessel from GDCS pools. This occurs ten
minutes after the initiation of accidents. The reactor core is never uncovered during the limiting
of DBA. The steam condensation in the suppression pool and pressure equilibrium between dry
well and wet well through the vacuum breakers reduces the dry well pressure causing the
horizontal vents to close. The remaining non-condensable gases and steam in dry well then
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follow up through the POCS heat exchanger. The steam is condensed as it passes through the
PCCS tubes. Water condensate is collected and returns to GDCS pools and the non condensable
gases flow into the wet well gas space .This establishes a passive long term recirculation cooling
mode for over 72 hours-non safety related recalculating fans are credited after 72 hours and
result in further reduction in containment pressure. However calculations show that even in
purely passive mode the containment pressure remains below design pressure for over 30 days.
ESBWR is the Largest Capacity Model
Summary
Country &
Vendor
Reactor
Type
C1 C2 C3 C4 C5 C6
USA-
JAPAN GE-
Hitachi
ESBWR
JAPAN
Mitsubishi
US-
APWR
KOREA
KEPCO
Consortium
APR-
1400
6.5 Russia Gidopress
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Reactor Type: VVER-1200 (PWR)
Reactor Capacity: 1200 Mwe
The VVER, or WWER, (from Russian: Водо-водяной энергетический реактор; transliterates
as Vodo-Vodyanoi Energetichesky Reactor; Water-Water Power Reactor) is a series of
pressurised water reactor designs originally developed in the Soviet Union, and now Russia, by
OKB Gidropress. Power output ranges from 440 MWe to 1200 MWe with the latest Russian
development of the design. VVER power stations are used by Armenia, Bulgaria, China, Czech
Republic, Finland, Hungary, India, Iran, Slovakia, Ukraine, and the Russian Federation.
The Russian abbreviation VVER stands for water-cooled, water-moderated energy reactor. This
describes the pressurised water reactor (PWR) design. The main distinguishing features of the
VVER compared to other PWRs are:
Horizontal steam generators
Hexahedral fuel assemblies
No bottom penetrations in the pressure vessel
High-capacity pressurisers providing a large reactor coolant inventory
Reactor fuel rods are fully immersed in water kept at 15 MPa of pressure so that it does not boil
at normal (220 to over 300 °C) operating temperatures. Water in the reactor serves both as a
coolant and a moderator which is an important safety feature. Should coolant circulation fail the
neutron moderation effect of the water diminishes, reducing reaction intensity and compensating
for loss of cooling, a condition known as negative void coefficient. Later versions of the reactors
are encased in massive steel pressure shells. Fuel is low enriched (ca. 2.4–4.4% 235U) uranium
dioxide (UO2) or equivalent pressed into pellets and assembled into fuel rods.
Reactivity is controlled by control rods that can be inserted into the reactor from above. These
rods are made from a neutron absorbing material and depending on depth of insertion hinder the
chain reaction. If there is an emergency, a reactor shutdown can be performed by full insertion of
the control rods into the core.
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The VVER-1200 (or NPP-2006 or AES-2006) is an evolution of the VVER-1000 being offered
for domestic and export use. Specifications include a $1,200 per kW electric capital cost, 54
month planned construction time, and expected 50 year lifetime at 90% capacity factor. The
VVER 1200 will produce 1,200 MWe of power. Safety features include a containment building
and missile shield. It will have full emergency systems that include an emergency core cooling
system, emergency backup diesel power supply, advanced refuelling machine, computerized
reactor control systems, backup feedwater supply and reactor SCRAM system. The nuclear
reactor and associated systems will be hosted in one single building and there will be another
building for the turbo generators. The main building will comprise the reactor, refueling machine
and diesel backup power supply, steam generators and reactor control systems.
If a VVER-1200 experiences a loss of coolant accident or loss of power accident the turbo
generators 'coast down' for 30 seconds, during which time a shutdown can be initiated using
residual power in the system. Further emergency power is available from a backup set of diesel
generators kept on standby to maintain cooling flow to the reactor. The reactor design has been
refined to optimize fuel efficiency.
The first two units are proposed for Leningrad Nuclear Power Plant II and Novovoronezh
Nuclear Power Plant II. A standardized design has not been elected. Mainly are more reactors
with a VVER-1200/491 like the Leningrad-II-design are firmly planned (Kaliningrad and Nizhny
Novgorod NPP) and under construction. The VVER-1200/392M under construction at the
Novovoronezh NPP-II is selected for the Seversk, Zentral and South-Urals NPP. A standard
version was developed as VVER-1200/510 and referred to as VVER-TOI.
6.6 USA Westinghouse
Reactor Type : AP-1000
Reactor Capacity : 1100 MWe
The AP1000 is a two-loop pressurized water reactor planned to produce a net power output of
1,117 MWe. It is an evolutionary improvement on the AP600, essentially a more powerful model
with roughly the same footprint.
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The design is less expensive to build than other Generation III designs partly because it uses
existing technology. The design also decreases the number of components, including pipes,
wires, and valves. Standardization and type-licensing should also help reduce the time and cost
of construction. Because of its simplified design compared to a Westinghouse generation II
PWR, the AP1000 has.
50% fewer safety-related valves
35% fewer pumps
80% less safety related piping
85% less control cable
45% less seismic building volume
The AP1000 design is considerably more compact in land usage than most existing PWRs, and
uses under a fifth of the concrete and rebar reinforcing of older designs.
Probabilistic risk assessment was used in the design of the plants. This enabled minimization of
risks, and calculation of the overall safety of the plant. According to the NRC, the plants will be
orders of magnitude safer than those in the last study, NUREG-1150. The AP1000 has a
maximum core damage frequency of 2.41 × 10−7 per plant per year.
Used fuel produced by the AP1000 can be stored indefinitely in water on the plant site. Aged
used fuel may also be stored in above-ground dry cask storage, in the same manner as the
currently operating fleet of U.S. power reactors.
Power reactors of this general type continue to produce heat from radioactive decay products
even after the main reaction is shut down, so it is necessary to remove this heat to avoid
meltdown of the reactor core. In the AP1000, Westinghouse's Passive Core Cooling System uses
multiple explosively-operated and DC operated valves which must operate within the first
30 minutes. This is designed to happen even if the reactor operators take no action. The electrical
system required for initiating the passive systems doesn't rely on external or diesel power and the
valves don't rely on hydraulic or compressed air systems. The design is intended to passively
remove heat for 72 hours, after which its gravity drain water tank must be topped up for as long
as cooling is required.
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US EPRI and ALWR URD
The AP1000 pressurized water reactor works on the simple concept that, in the event of a design-
basis accident (such as a coolant pipe break), the plant is designed to achieve and maintain safe
shutdown condition without any operator action and without the need for ac power or pumps.
Instead of relying on active components such as diesel generators and pumps, the AP1000 relies
on the natural forces of gravity, natural circulation and compressed gases to keep the core and
containment from overheating. However, many active components are included in the AP1000,
but are designated as non safety-related. Multiple levels of defense for accident mitigation are
provided, resulting in extremely low core-damage probabilities while minimizing occurrences of
containment flooding, pressurization and heat-up. The AP1000 is the safest and most economical
nuclear power plant available in the worldwide commercial marketplace, and is the only
Generation III+ reactor to receive Design Certification from the U.S. Nuclear Regulatory
Commission (NRC).
The reactor is proven and standardized
China has officially adopted the AP1000 as a standard for inland nuclear projects. The National
Development and Reform Commission (NDRC) has already approved several nuclear projects,
including the Dafan plant in Hubei province, Taohuajiang in Hunan, and Pengze in Jiangxi. The
NDRC is studying additional projects in Anhui, Jilin and Gansu provinces. China wants to have
100 units under construction and operating by 2020, according to Aris Candris, Westinghouse's
CEO. In 2008 and 2009 Westinghouse made agreements to work with the State Nuclear Power
Technology Corporation (SNPTC) and other institutes to develop a larger design, the CAP1400
of 1400 MWe capacity, possibly followed by a 1700 MWe design. China will own the
intellectual property rights for these larger designs. Exporting the new larger units may be
possible with Westinghouse's cooperation.
6.7 France Areva
Reactor Type : NP EPR
Reactor Capacity : 1600
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The main design objectives of the generation III EPR design are increased safety while providing
enhanced economic competitiveness through improvements to previous PWR designs scaled up
to an electrical power output of around 1650 MWe (net) with thermal power 4500 MWt. The
reactor can use 5% enriched uranium oxide fuel, reprocessed uranium fuel and 100% mixed
uranium plutonium oxide fuel. The EPR is the evolutionary descendant of the Framatome N4
and Siemens Power Generation Division KONVOI reactors.
The EPR design has several active and passive protection measures against accidents:
Four independent emergency cooling systems, providing the required cooling of the
decay heat that continues for 1 to 3 years after the reactor's initial shutdown (i.e. 300%
redundancy)
Leak tight containment around the reactor
An extra container and cooling area if a molten core manages to escape the reactor (see
containment building)
Two-layer concrete wall with total thickness 2.6 meters, designed to withstand impact by
aeroplanes and internal overpressure
The reactor is proven and standardized
The construction of the Olkiluoto 3 power plant in Finland commenced in August 2005. It was
initially scheduled to go online in 2009, but the project has suffered many delays, and operation
is now expected to start in 2014. It is still expected to be the first EPR reactor built in the world.
The plant will have an electrical power output of 1600 MWe (net). The construction is a joint
effort of French Areva and German Siemens AG through their common subsidiary Areva NP, for
Finnish operator TVO. Initial cost estimates were about € 3.7 billion,
Use of enriched Uranium Fuels
AREVA offers an array of customizable services spanning the entire uranium life cycle: from
mines to the manufacture of fuel assemblies, including ore conversion and enrichment.The range
of uranium-related services enables AREVA's clients to benefit from the competence,
technologies, reliability and industrial capacity of the group in any of its fields of expertise:
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Mines
o Reliable supply thanks to diversified and large-scale resources (more than
200,000 tons of reserves), a portfolio of mining projects already underway and
varied exploration activities.
Conversion
o Guarantee of a high-quality product ready for enrichment using the most
advanced techniques.
Enrichment
o The performance of the centrifuging enrichment process adds to AREVA's
experience in the field.
Fuel design and fabrication
o Optimized use of fuel under increasingly demanding operating conditions.
Power rating Greater Than 500 MWe
The EPR reactor has an electrical production capacity of more than 1650 MWe, which places it
among the most powerful reactors in the world. A direct descendant of previous models
manufactured by AREVA, the EPR pressurized water reactor is based on tried-and-tested
technologies and principles. It is classified as a generation III+ reactor due to the level of safety
obtained and the economic savings that it achieves in relation to the earlier models.
6.8 CANADA AECL
Reactor Type : ACR-1080
Reactor Capacity : 1080 MWe
The advanced CANDU reactor (ACR) is a Generation III+ nuclear reactor design and is a
further development of existing CANDU reactors designed by Atomic Energy of Canada
Limited. The ACR is a light-water-cooled reactor that incorporates features of both Pressurised
Heavy Water Reactors (PHWR) and advanced pressurized water reactors (APWR) technologies.
It uses a similar design concept to the steam-generating heavy water reactor (SGHWR).
The design uses lightly enriched uranium (LEU) fuel, ordinary (light) water coolant, and a
separate heavy water moderator. The reactivity regulating and safety devices are located within
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the low pressure moderator. The ACR also incorporates characteristics of the CANDU design,
including on-power refueling with the CANFLEX fuel; a long prompt neutron lifetime; small
reactivity holdup; two fast, totally independent, dedicated safety shutdown systems; and an
emergency core cooling system (although all generation 2, 3, and 3+ designs have this feature).
The compact reactor core reduces core size by half for the same power output over the older
design.
The fuel bundle is a variant of the 43-element CANFLEX design (CANFLEX-ACR). The use of
LEU fuel with a neutron absorbing centre element allows the reduction of coolant void reactivity
coefficient to a nominally small, negative value. It also results in higher burnup operation than
traditional CANDU designs.
Safety
The ACR-1080 design currently calls for a variety of safety systems, most of which are
evolutionary derivatives of the systems utilized on the CANDU 6 reactor design. Each ACR
requires both SDS1 & SDS2 to be online and fully operational before they will operate at any
power level.
Safety Shutdown System 1 (SDS1): SDS1 is designed to rapidly and automatically terminate
reactor operation. Neutron-absorbing rods (control rods that shut down the nuclear chain
reaction) are stored inside isolated channels located directly above the reactor vessel (Calandria)
and are controlled via a triple-channel logic circuit. When any 2 of the 3 circuit paths are
activated (due to sensing the need for emergency reactor trip), the direct current-controlled
clutches that keep each control-rod in the storage position are de-energized. The result is that
each control-rod is inserted into the Calandria, and the reactor heat output is reduced by 90%
within 2 seconds.
Safety Shutdown System 2 (SDS2): SDS2 is also designed to rapidly and automatically
terminate reactor operation. Gadolinium nitrate (GdNO3) solution, a neutron-absorbing liquid
that shuts down the nuclear chain reaction, is stored inside channels that feed into horizontal
nozzle assemblies. Each nozzle has an electronically controlled valve, all of which are controlled
via a triple-channel logic circuit. When any 2 of the 3 circuit paths are activated (due to sensing
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the need for emergency reactor trip), each of these valves are opened and liquid GdNO3 is
injected through the nozzles to mix with the heavy-water moderator liquid in the reactor vessel
(Calandria). The result is that the reactor heat output is reduced by 90% within 2 seconds.
Reserve water system (RWS): The RWS consists of a water tank located at a high elevation
within the reactor building. This provides water for use in cooling an ACR that has suffered a
Loss of Coolant Accident (LOCA). The RWS can also provide emergency water (via gravity-
feed) to the steam generators, moderator system, shield cooling system or the heat transport
system of any ACR.
Electrical power supply system (EPS): The EPS system is designed to provide each ACR unit
with the required electrical power needed to perform all safety functions under both operating &
accident conditions. It contains seismically qualified, redundant standby generators, batteries and
distribution switchgear.
Cooling water system (CWS): The CWS provides all necessary light water (H2O) required to
perform all safety system-related functions under both operating & accident conditions. All
safety-related portions of the system are seismically qualified and contain redundant divisions
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7.0 Summary of Reactor Technologies
Country &
VendorReactor Type
Reactor
Capacity
(MWe)
Remarks
USA
WestinghouseAP-1000 (PWR) 1100
AP-1000NRC certification 2005, first units
being built in China, many more planned.
Simplified construction and operation.
3 years to build.
60-year plant life.
France Areva NP EPR (PWR) 1600
French design approval. Being built in Finland
and France, planned for China. US version
developed.
Evolutionary design.
High fuel efficiency.
Flexible operation.
USA-Japan
GE-HitachiESBWR (BWR) 1550
Developed from ABWR, under certification in
USA, likely construction there.
Evolutionary design.
Short construction time.
Japan
Mitsubishi
US-APWR
(PWR)1700
Basic design in progress, planned for Tsuruga
US DC application 2008.
Hybrid safety features.
Simplified Construction and operation.
Korea KEPCO
Consortium
APR-1400
(PWR)1450
Design certification 2002, First units expected
to be operating in 2013.
Evolutionary design.
Increased reliability.
Simplified construction and operation.
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Russia
Gidropress
VVER-1200
(PWR)1200
Replacement under construction for Leningrad
and Novovoronezh plants.
Evolutionary design.
High fuel efficiency.
50-year plant life
Canada AECLACR-1080
(PHWR)1080
Undergoing certification in Canada
Evolutionary design.
Light water cooling.
Low-enriched fuel.
8.0 Selected Reactor Technologies.
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