Nuke Edited 1_1

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.

17


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

18


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

19


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.

20


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

21


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.

22


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

23


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.

24


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

25


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

26


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

27


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

28


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.

29


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.

30


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.

31


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

32


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:

33


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

34


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.


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.


Increased reliability.

Simplified construction and operation.

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Russia

Gidropress

VVER-1200

(PWR)1200

Replacement under construction for Leningrad

and Novovoronezh plants.


High fuel efficiency.

50-year plant life

Canada AECLACR-1080

(PHWR)1080

Undergoing certification in Canada


Light water cooling.

Low-enriched fuel.

8.0 Selected Reactor Technologies.

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