Download pdf - Chapter 07_Generation III Advanced Nuclear Reactors - Part2

8/3/2019 Chapter 07_Generation III Advanced Nuclear Reactors - Part2

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Chapter07:Generation III Advanced Nuclear Reactors Part2

Edited byDr. Mir F. Ali1

Chapter07GenerationIIIAdvancedNuclear

Reactors

Light Water Reactors (LWR) was, covered in previous chapter, and the remaining types ofnuclear power reactorsHeavy Water Reactors (HWR), High Temperature Gas-Cooled

Reactors (HTGR), and Fast Neutron Reactors (FNR) will be covered under this chapter.

No Description

1. LGHT WATER REACTORS (LWR):

1.1 European Pressurized Water Reactor (EPR);

1.2 Advanced Passive 1000 (AP1000);

1.3 Advanced Boiling Water Reactors (ABWR);

1.4 Economic & Simplified Boiling Water Reactors (ESBWR);

1.5 Advanced Pressurized Water Reactors (APWR);

1.6 Advanced Pressurized Reactors 1400 (APR1400);

1.7 Atmea1;

1.8 Kerena/Karena;1.9 AES-92,V392;

1.10 AES-2006;

1.11 MIR-1200;

1.12 International Reactor Innovation and Secure (IRIS);

1.13 VBER-300; and

2. HEAVY WATER REACTORS (HWR):

2.1 Enhanced CANDU-6(EC6);

2.2 Advanced CANDU Reactors (ACR);

2.3 Advanced Heavy Water Reactors (AHWR);

3. HIGH TEMPERATURE GAS-COOLD REACTORS (HTGR):

3.1 HTR-PM;3.2 Pebble Bed Modular Reactors (PBMR); and

3.3 Gas TurbineModular Helium Reactors (G7-MHR).

4. FAST NEUTRON REACTORS (FNR):

4.1 Fast Breeder Reactors (FBR);

4.2 Japan Standard Fast Reactors (JSFR);

4.3 BN-600;

4.4 BN-800;

4.5 BREST;

4.6 European Lead-Cooled System (ELSY);

4.7 PRISM; and

4.8 KALIMER.

2. HEAVY WATER REACTORS (HWR):Heavy water reactors use heavy water as a neutron moderator. Heavy water is deuteriumoxide, D2O. Neutrons in a nuclear reactor that uses uranium must be slowed down sothat they are more likely to split other atoms and get more neutrons released to splitother atoms.


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The HWR concept allow the use of natural uranium as a fuel without the need for itsenrichment, offering a degree of energy independence, especially if uranium is availablefor mining or for extraction as a byproduct of another industry such as gold mining orphosphate fertilizer production. However, it needs the installation of a heavy waterD2Oproduction capability which is a much simpler endeavor anyway since separating thelight Isotopes (D from H) is much simpler than separating the heavy Isotopes.

HWR have become a significant proportion of world reactor installations second only tothe Light Water Reactors (LWR). They provide fuel cycle flexibility for the future and canpotentially burn the recycled fuel from LWR, with no major reactor design changes, thusextending resources and reducing spent fuel storage.

In Canada, the government-owned Atomic Energy of Canada Ltd (AECL) has had twodesigns under development, which are, based on its reliable CANDU-6 reactors, the mostrecent of which are operating in China.

The CANDU-9 (925-1300 MWe) was, developed from this also as a single-unit plant. It hasflexible fuel requirements ranging from natural uranium through slightly enricheduranium, recovered uranium from reprocessing spent PWR fuel, mixed oxide (U & Pu)fuel, direct use of spent PWR fuel, to thorium. It may be able to burn military plutoniumor actinides separated from reprocessed PWR/BWR waste. A two year licensing review ofthe CANDU-9 design was successfully completed early in 1997, but the design has beenshelved.

Here is a brief description on each reactor that is defined under this category:

2.1 Enhanced CANDU-6 (EC6):Some of the innovation of this along with experience in building recent Korean andChinese units was then put back into the Enhanced CANDU-6 (EC6) built as twin units with power increase to 750 MWe gross (690 MWe net, 2084 MWt) and flexible fueloptions, plus 4.5 year construction and 60-year plant life (with mid-life pressure tubereplacement). This is under consideration for new build in Ontario. AECL claims it as aGeneration III design.

The Advanced CANDU Reactor (ACR), a 3rd generation reactor and is a more innovativeconcept. While retaining the low-pressure heavy water moderator, it incorporates somefeatures of the pressurized water reactor. Adopting light water cooling and a more

compact core reduces capital cost and because the reactor is run at higher temperatureand coolant pressure it has higher thermal efficiency.


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2.2 Advanced CANDU Reactor (ACR):The ACR-700 design was 700 MWe but is physically much smaller, simpler and moreefficient as well as 40 percent cheaper than the CANDU-6. However, the ACR-1000 of1080-1200 MWe (3200 MWt) is now the focus of attention byAECL.
http://www.aecl.ca/index.asphttp://www.aecl.ca/index.asp


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It has more fuel channels (each of which can be regarded as a module of about 2.5 MWe).The ACR will run on low-enriched uranium (about 1.5-2.0 percent U-235) with high burn-up, extending the fuel life by about three times and reducing high-level waste volumesaccordingly. It will also efficiently burn MOX fuel, thorium and actinides.

Regulatory confidence in safety is enhanced by a small negative void reactivity for the

first time in CANDU and utilizing other passive safety features as well as two independentand fast shutdown systems. Units will be assembled from prefabricated modules cuttingconstruction time to 3.5 years. ACR units can be built singly but are optimal in pairs. Theywill have 60 year design life overall but require mid-life pressure tube replacement.

ACR is moving towards design certification in Canada with a view to following in China,USA and UK. In 2007, AECL applied for UK generic design assessment (pre-licensingapproval) but then withdrew after the first stage. In the USA, NRC lists the ACR-700 asbeing at pre application review stage. The first ACR-1000 unit could be operating in 2016in Ontario.

The CANDU Xor SCWR is a variant of the ACR, but with supercritical light water coolant(eg 25 MPa and 625C) to provide 40 percent thermal efficiency. The size range envisagedis 350 to 1150 MWe depending on the number of fuel channels used. Commercializationenvisaged after 2020.

2.3 Advanced Heavy Water Reactor (AHWR):India is developing the Advanced Heavy Water reactor (AHWR) as the third stage in itsplan to utilize thorium to fuel its overall nuclear power program. The AHWR is a 300MWe gross (284 MWe net, 920 MWt) reactor moderated by heavy water at low pressure.


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The calandria has about 450 vertical pressure tubes and the coolant is boiling light watercirculated by convection. A large heat sink Gravity-driven water pool with 7000cubic metres of water is near the top of the reactor building. Each fuel assembly has 30Th-U-233 oxide pins and 24 Pu-Th oxide pins around a central rod with burnableabsorber. Burn-up of 24 GWd/t is, envisaged. It is designed to be self-sustaining inrelation to U-233 bred from Th-232 and have a low Pu inventory and consumption with

slightly negative void coefficient of reactivity. It is designed for 100-year plant life and isexpected to utilize 65 percent of the energy of the fuel with two thirds of that energycoming from thorium via U-233.

Once it is operational each AHWR fuel assembly will have the fuel pins arranged in threeconcentric rings arranged:

Inner: 12 pins Th-U-233 with 3.0 percent U-233; Intermediate: 18 pins Th-U-233 with 3.75 percent U-233; and Outer: 24 pins Th-Pu-239 with 3.25 percent Pu.

The fissile plutonium content will decrease from an initial 75 percent to 25 percent atequilibrium discharge burn-up level. As well as U-233 some U-232 is formed and thehighly gamma-active daughter products of this confer a substantial proliferationresistance.

In 2009, an export version of this design was announced: theAHWR-LEU. This will uselow-enriched uranium plus thorium as fuel dispensing with plutonium input. About 39percent of the power will come from thorium (via in situ conversion to U-233) and burn-up will be 64 GWd/t. Uranium enrichment level will be 19.75 percent giving 4.21 percentaverage fissile content of the U-Th fuel. While designed for closed fuel cycle this is notrequired. Plutonium production will be less than in light water reactors and the fissile

proportion will be less and the Pu-238 portion three times as high giving inherentproliferation resistance. The AEC says, The reactor is manageable with modest industrialinfrastructure within the reach of developing countries. In the AHWR-LEU the fuelassemblies will be configured:

Inner ring: 12 pins Th-U with 3.555 percent U-235; Intermediate ring: 18 pins Th-U with 4.345 percent U-235; and Outer ring: 24 pins Th-U with 4.444 percent U-235.

3. HIGH-TEMPERATURE GAS-COOLED REACTORS (HTGR):High Temperature Gas cooled Reactors (HTGR) distinguished from other gas-cooled

reactors by the higher temperatures attained within the reactor. Such highertemperatures might permit the reactor to be used as an industrial heat source in additionto generating electricity.

Among the future uses for which HTGR are being considered is the commercialgeneration of hydrogen from water. In some cases HTGR turbines run directly by the gasthat is used as a coolant. In other cases steam or alternative hot gases such as nitrogen are


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produced in a heat exchanger to run the power generators. Recent proposals havefavoured helium as the gas used as an HTGR coolant.

These reactors use helium as a coolant at up to 950C which either makes steamconventionally or directly drives a gas turbine for electricity and a compressor to returnthe gas to the reactor core. Fuel is in the form of TRISO particles less than a millimetre in

diameter. Each has a kernel of uranium oxycarbide with the uranium enriched up to 17percent U-235. Layers of carbon and silicon carbide giving containment for fissionproducts which is stable to 1600C or more surround this. These particles may bearranged: in blocks as hexagonal prisms of graphite, or in billiard ball-sized pebbles ofgraphite encased in silicon carbide.

Here is a brief description for each reactor under this category:

3.1 HTR-PM:The first commercial version will be Chinas HTR-PM being built at Shidaowan inShandong province. It has been developed by Tsinghua Universitys INET which is the

R&D leader and Chinergy Co., with China Huaneng Group leading the demonstrationplant project.

This reactor will have two reactor modules, each of 250 MWt/ 105 MWe; using 9 percentenriched fuel (520,000 elements) giving 80 GWd/t discharge burnup. With an outlettemperature of 750 C, the pair will drive a single steam cycle turbine at about 40 percentthermal efficiency. This 210 MWe Shidaowan demonstration plant is to pave the way foran 18-unit (3x6x210MWe) full-scale power plant on the same site, also using the steamcycle. The plant life is envisaged as 60 years with 85 percent load factor.

3.2 Pebble Bed Modular Reactor (PBMR):


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South Africas PBMR was being developed by a consortium led by the utility Eskom withMitsubishi Heavy Industries from 2010.

It draws on German expertise. It aims for a step change in safety, economics andproliferation resistance. Production units would be 165 MWe. The PBMR will ultimatelyhave a direct-cycle (Brayton cycle) gas turbine generator and thermal efficiency about 41percent the helium coolant leaving the bottom of the core at about 900C and driving aturbine. Power is adjusted by changing the pressure in the system. The helium is passedthrough a water-cooled pre-cooler and intercooler before being returned to the reactorvessel. (In the Demonstration Plant, it will transfer heat in a steam generator rather thandriving a turbine directly.)

Up to 450,000 fuel pebbles recycle through the reactor continuously (about six timeseach) until they are expended giving an average enrichment in the fuel load of 4-5 percentand average burn-up of 80 GWday/t U (eventual target burn-ups are 200 GWd/t). Thismeans on-line refueling as expended pebbles is replaced giving high capacity factor. Eachunit will finally discharge about 19 tonnes/yr of spent pebbles to ventilated on-site storagebins. A reactor will use about 13 fuel loads in a 40-year lifetime. Operational cycles areexpected to be six years between shutdowns.

Performance includes great flexibility in loads (40-100 percent) with rapid change inpower settings. Power density in the core is about one tenth of that in a light waterreactor and if coolant circulation ceases the fuel will survive initial high temperatureswhile the reactor shuts itself down giving inherent safety. Overnight capital cost (whenin clusters of eight units) is expected to be modest and generating cost very competitive.However, development has ceased due to lack of funds and customers.

3.3 Gas Turbine Modular Helium Reactor (GT-MHR):


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A larger US design the GT-MHR, is, planned as modules of 285 MWe each directly drivinga gas turbine at 48 percent thermal efficiency.

The cylindrical core consists of 102 hexagonal fuel element columns of graphite blockswith channels for helium and control rods. Graphite reflector blocks are both inside andaround the core. Half the core is, replaced every 18 months. Burn-up is about 100,000MWd/t. It is being developed by General Atomics in partnership with Russias OKBMAfrikantov supported by Fuji (Japan). Initially it was to be used to burn pure ex-weaponsplutonium at Seversk (Tomsk) in Russia. The preliminary design stage was completed in2001, but the program has stalled since. Arevas Antares is based on the GT-MHR.

4. FAST NEUTRON REACTORS:Fast neutron reactors are a technological step beyond conventional power reactors. Theyoffer the prospect of vastly more efficient use of uranium resources and the ability to burnactinides which are otherwise the long-lived component of high-level nuclear wastes.Some 300 reactor-years experience has been, gained in operating them

About 20 Fast Neutron Reactors (FNR) have already been operating some since the 1950s, and

some supply electricity commercially. Over 300 reactor-years of operating experience have been

accumulated. These more deliberately use the uranium-238 as well as the fissile U-235 isotopeused in most reactors. If they are designed to produce more plutonium than they consume theyare called Fast Breeder Reactors (FBR). If they are net consumers of plutonium they are

sometimes called burners. Fast neutron reactors also can burn long-lived actinides which are

recovered from used fuel out of ordinary reactors.

Several countries have research and development programs for improved Fast NeutronReactors and the IAEAs INPRO program involving 22 countries (see later section) has fast


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neutron reactors as a major emphasis in connection with closed fuel cycle. For instance,one scenario in France is for half of the present nuclear capacity to be replaced by fastneutron reactors by 2050 (the first half being replaced by 3rd-generation EPR units).

The FNR was originally conceived to burn uranium more efficiently and thus extend theworlds uranium resources it could do this by a factor of about 60. When thoseresources were perceived to be scarce several countries embarked upon extensive FBRdevelopment programs. However, significant technical and materials problems wereencountered and geological exploration showed by the 1970s that uranium scarcity wasnot going to be a concern for some time. Due to both factors by the 1980s it was clear thatFNRs would not be commercially competitive with existing light water reactors for sometime. Here are some Fast Neutron Reactors:

4.1 Fast Breeder Reactor (FBR):Several countries have research and development programs for improved Fast BreederReactors (FBR), which are a type of Fast Neutron Reactor. These use the uranium-238 in

reactor fuel as well as the fissile U-235 isotope used in most reactors.About 20 liquid metal-cooled FBR have already been operating, some since the 1950s, andsome have supplied electricity commercially. About 300 reactor-years of operatingexperience have been, accumulated.


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Natural uranium contains about 0.7 percent U-235 and 99.3 percent U-238. In any reactorthe U-238 component is turned into several isotopes of plutonium during its operation.Two of these Pu 239 and Pu 241 then undergo fission in the same way as U 235 to produceheat. In a fast neutron reactor this process is optimized so that it can breed fuel, oftenusing a depleted uranium blanket around the core. FBR can utilize uranium at least 60times more efficiently than a normal reactor. They are however expensive to build andcould only be justified economically if uranium prices were to rise to pre-1980 values wellabove the current market price. For this reason research work almost ceased for someyears and that on the 1450 MWe European FBR has apparently lapsed. Closure of the 1250MWe French Superphenix FBR after very little operation over 13 years also set backdevelopments.

Research continues in India at the Indira Gandhi Centre for Atomic Research a 40 MWtfast breeder test reactor has been operating since 1985. In addition, the tiny Kamini thereis employed to explore the use of thorium as nuclear fuel by breeding fissile U-233. In2004 construction of a 500 MWe prototype fast breeder reactor started at Kalpakkam.

The unit is expected to be operating in 2011 fuelled with uranium-plutonium carbide (thereactor-grade Pu being from its existing PHWR) and with a thorium blanket to breedfissile U-233. This will take Indias ambitious thorium program to stage two and set thescene for eventual full utilization of the countrys abundant thorium to fuel reactors.

Japan plans to develop FBR and its Joyo experimental reactor which has been operatingsince 1977 is now being boosted to 140 MWt. The 280 MWe Monju prototypes


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commercial FBR was connected to the grid in 1995, but was then shut down due to asodium leak. Its restart is, planned for 2009.

4.2 Japan Standard Fast Reactor (JSFR):Mitsubishi Heavy Industries (MHI) is involved with a consortium to build the JapanStandard Fast Reactor (JSFR) concept though with breeding ratio less than 1:1. This is alarge unit which will burn actinides with uranium and plutonium in oxide fuel. It could

be of any size from 500 to 1500 MWe. In this connection, MHI has also set up MitsubishiFBR Systems (MFBR).

4.3 BN-600:The Russian BN-600 fast breeder reactor at Beloyarsk has been supplying electricity to thegrid since 1981 and has the best operating and production record of all Russias nuclearpower units.

It uses uranium oxide fuel and the sodium coolant delivers 550C at little more thanatmospheric pressure. The BN 350 FBR operated in Kazakhstan for 27 years and about

half of its output was used for water desalination. Russia plans to reconfigure the BN-600to burn the plutonium from its military stockpiles.

4.4 BN-800:The first BN-800 a new larger (880 MWe) FBR from OKBM with improved features isbeing built at Beloyarsk. It has considerable fuel flexibility U+Pu nitride, MOX, or metaland with breeding ratio up to 1.3. It has much enhanced safety and improved economyoperating cost is expected to be only 15 percent more than VVER. It is capable of burning


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2 tonnes of plutonium per year from dismantled weapons and will test the recycling ofminor actinides in the fuel. The BN-800 has been sold to China and two units are due tostart construction there in 2012.

However, the Beloyarsk-4 BN-800 is likely to be the last such reactor built (outside Indiasthorium program), with a fertile blanket of depleted uranium around the core. Furtherfast reactors will have an integrated core to minimize the potential for weaponsproliferation from bred Pu-239. Beloyarsk-5 is, designated as a BREST design.

4.5 BREST:Russia has experimented with several lead-cooled reactor designs and has used lead-bismuth cooling for 40 years in reactors for its seven Alfa class submarines.


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Pb-208 (54 percent of naturally occurring lead) is transparent to neutrons. A significantnew Russian design from NIKIET is the BREST fast neutron reactor, of 300 MWe or morewith lead as the primary coolant at 540C, and supercritical steam generators. It isinherently safe and uses a high-density U+Pu nitride fuel with no requirement for highenrichment levels. No weapons-grade plutonium can be produced (since there is nouranium blanket all the breeding occurs in the core).

In addition, it is an equilibrium core so there are no spare neutrons to irradiate targets.The initial cores can comprise Pu and spent fuel hence loaded with fission products andradiologically hot. Subsequently, any surplus plutonium which is not in pure form canbe, used as the cores of new reactors. Used fuel can be recycled indefinitely with on-sitereprocessing and associated facilities. A pilot unit is planned for Beloyarsk by 2020 and1200 MWe units are proposed.

4.6 European Lead-Cooled System (ELSY):

The European Lead-cooled System (ELSY) of 600 MWe in Europe led by Ansaldo Nuclearfrom Italy and financed by Erratum. ELSY is a flexible fast neutron reactor which can usedepleted uranium or thorium fuel matrices and burn actinides from LWR fuel. Liquidmetal (Pb or Pb-Bi eutectic) cooling is at low pressure. The design was nearly complete in2008 and a small-scale demonstration facility is planned. It runs on MOX fuel at 480Cand the molten lead is pumped to eight steam generators though decay heat removal ispassive by convection.

4.7 PRISM:


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In the USA GE was involved in designing a modular liquid metal-cooled inherently safereactor PRISM. GE with the DOE national laboratories were developing PRISM duringthe advanced liquid-metal fast breeder reactor (ALMR) program. No US fast neutronreactor has so far been larger than 66 MWe and none has supplied electricitycommercially.

Todays PRISM is a GE-Hitachi design for compact modular pool-type reactors withpassive cooling for decay heat removal. After 30 years of development it represents GEHsGeneration IV solution to closing the fuel cycle in the USA. Each PRISM Power Blockconsists of two modules of 311 MWe each, operating at high temperature over 500C.

The pool-type modules below ground level contain the complete primary system withsodium coolant. The Pu & DU fuel is metal and obtained from used light water reactorfuel. However, all transuranic elements are, removed together in the electrometallurgical

reprocessing so that fresh fuel has minor actinides with the plutonium. Fuel stays in thereactor about six years, with one third removed every two years. Used PRISM fuel isrecycled after removal of fission products. The commercial-scale plant concept part of anAdvanced Recycling Centre uses three power blocks (six reactor modules) to provide 1866MWe.

4.8 KALIMER (KAERI):Koreas KALIMER (Korea Advanced Liquid Metal Reactor) is a 600 MWe pool typesodium-cooled fast reactor designed to operate at over 500C. It has evolved from a 150


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MWe version. It has a transmuter core and no breeding blanket is involved. Futuredevelopment of KALIMER as a Generation IV type is envisaged.

Advanced water-cooled-reactor nuclear energy system concepts have been identified aspart of the Generation IV International Roadmap evaluation and R&D planning activity;i.e., involving international laboratories, academia, and industry groups from countries

including Argentina, Brazil, Canada, France, Italy, Japan, Korea, Russia, Switzerland, theUK and the U.S. This activity resulted in the proposal of over thirty-eight specific reactordesigns. The leading reactor designs can be, categorized into two general groups:

1. Near-Term Advanced Boiling Water and Pressurized Water Reactors with Passive-Safety; and

2. Longer-Term Advanced Water Reactors e.g., Supercritical Water Reactor.The first grouping of advanced Boiling Water Reactor (BWR) and Pressurized WaterReactors (PWR) systems can be represented by the Experimental Simplified BoilingWater Reactor (ESBWR)) and the Advanced Pressurized Water Reactor (AP1000), whilethe Supercritical Water Reactor (SCWR) is a unique example of the second grouping.

Advanced reactors have also been, proposed that utilize different coolants than water andpotentially may allow for more flexibility in operation, improved sustainability andminimizing by-product flows as well as providing the potential for higher outlettemperatures to allow for a wider range of process heat applications; e.g., high-temperature chemical reduction of water to produce hydrogen. Over fifty differentconcepts have been, proposed and the most promising designs can be, grouped into thefollowing:

1. Advanced Gas-Cooled Reactors for High Temperatures (PBMR, MHGR, VHTR,GFR); and

2. Advanced Liquid-Metal Fast Reactors (Sodium-cooled and Lead-alloy-cooled).The first grouping of advanced gas-cooled reactors can be represented by the Very HighTemperature Gas Reactor (VHTR) either with graphite pebbles or with prismatic graphiteblocks as moderators.

The second grouping can be represented by the integral sodium-cooled fast reactor or thelead-cooled fast reactor both providing high-temperature process heat with a low-pressure cooling circuit.

It is important to understand that while the third generation plants have been verysuccessful where they have been, built in Europe, Asia and the Pacific Rim, furtherevolution is needed to make new nuclear energy systems a more attractive option fordeployment around the world. In particular, the next generation of nuclear energysystems must be able to be licensed, constructed, and operated in a manner that willprovide a competitively priced supply of energy while satisfactorily addressing plantreliability, nuclear safety, waste disposal, proliferation resistance, and public perceptionconcerns of the countries in which they are deployed.


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This chapterwas published on Inuitech Intuitech Technologies for Sustainability onMay 20, 2011:http://intuitech.biz/?p=9432

Resources:

1. World Nuclear Association:http://www.world-nuclear.org/info/default.aspx?id=528&terms=Advanced percent20nuclearpercent20reactors

2. Advanced Nuclear Energy Systems:http://web.mit.edu/hmtl/www/papers/CORRADINI.pdf

3. GE Hitachi Nuclear Energy:http://www.gepower.com/prod_serv/products/nuclear_energy/en/downloads/gea14576e_abwr.pdf

4. Wikipedia:http://en.wikipedia.org/wiki/Economic_Simplified_Boiling_Water_Reactor

5. Atmea:http://www.atmea-sas.com/scripts/ATMEA/publigen/content/templates/show.asp?P=163&L=EN

6. Multiphase Flow Dynamics 4:http://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhCoq-sQOU7uj0Cw&sa=X&oi=book_result&ct=result&resnum=1&ved=0CBgQ6AEwAA#v=onepage&q=karena percent20reactor&f=false

7. AES-92 for Belene:http://www.ecology.at/files/pr529_1.pdf8. MIR-1200:http://www.atomstroyexport.ru/show/press/4699. Wikipedia:

http://en.wikipedia.org/wiki/International_Reactor_Innovative_and_Secure10.Heavy Water Reactors:https://netfiles.uiuc.edu/mragheb/www/NPRE

percent20402 percent20ME percent20405 percent20Nuclear percent20Powerpercent20Engineering/Heavy percent20Water percent20Reactor.pdf

11. Fast Neutron Reactors:http://www.vaec.gov.vn/userfiles/file/Fastpercent20Neutron percent20Reactors.pdf
http://intuitech.biz/?p=9432http://intuitech.biz/?p=9432http://intuitech.biz/?p=9432http://www.world-nuclear.org/info/default.aspx?id=528&terms=Advanced%20nuclear%20reactorshttp://www.world-nuclear.org/info/default.aspx?id=528&terms=Advanced%20nuclear%20reactorshttp://www.world-nuclear.org/info/default.aspx?id=528&terms=Advanced%20nuclear%20reactorshttp://www.world-nuclear.org/info/default.aspx?id=528&terms=Advanced%20nuclear%20reactorshttp://www.world-nuclear.org/info/default.aspx?id=528&terms=Advanced%20nuclear%20reactorshttp://web.mit.edu/hmtl/www/papers/CORRADINI.pdfhttp://web.mit.edu/hmtl/www/papers/CORRADINI.pdfhttp://www.gepower.com/prod_serv/products/nuclear_energy/en/downloads/gea14576e_abwr.pdfhttp://www.gepower.com/prod_serv/products/nuclear_energy/en/downloads/gea14576e_abwr.pdfhttp://www.gepower.com/prod_serv/products/nuclear_energy/en/downloads/gea14576e_abwr.pdfhttp://en.wikipedia.org/wiki/Economic_Simplified_Boiling_Water_Reactorhttp://en.wikipedia.org/wiki/Economic_Simplified_Boiling_Water_Reactorhttp://www.atmea-sas.com/scripts/ATMEA/publigen/content/templates/show.asp?P=163&L=ENhttp://www.atmea-sas.com/scripts/ATMEA/publigen/content/templates/show.asp?P=163&L=ENhttp://www.atmea-sas.com/scripts/ATMEA/publigen/content/templates/show.asp?P=163&L=ENhttp://www.atmea-sas.com/scripts/ATMEA/publigen/content/templates/show.asp?P=163&L=ENhttp://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhCoq-sQOU7uj0Cw&sa=X&oi=book_result&ct=result&resnum=1&ved=0CBgQ6AEwAA#v=onepage&q=karena%20reactor&f=falsehttp://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhCoq-sQOU7uj0Cw&sa=X&oi=book_result&ct=result&resnum=1&ved=0CBgQ6AEwAA#v=onepage&q=karena%20reactor&f=falsehttp://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhCoq-sQOU7uj0Cw&sa=X&oi=book_result&ct=result&resnum=1&ved=0CBgQ6AEwAA#v=onepage&q=karena%20reactor&f=falsehttp://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhCoq-sQOU7uj0Cw&sa=X&oi=book_result&ct=result&resnum=1&ved=0CBgQ6AEwAA#v=onepage&q=karena%20reactor&f=falsehttp://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhCoq-sQOU7uj0Cw&sa=X&oi=book_result&ct=result&resnum=1&ved=0CBgQ6AEwAA#v=onepage&q=karena%20reactor&f=falsehttp://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhCoq-sQOU7uj0Cw&sa=X&oi=book_result&ct=result&resnum=1&ved=0CBgQ6AEwAA#v=onepage&q=karena%20reactor&f=falsehttp://www.ecology.at/files/pr529_1.pdfhttp://www.ecology.at/files/pr529_1.pdfhttp://www.ecology.at/files/pr529_1.pdfhttp://www.atomstroyexport.ru/show/press/469http://www.atomstroyexport.ru/show/press/469http://www.atomstroyexport.ru/show/press/469http://en.wikipedia.org/wiki/International_Reactor_Innovative_and_Securehttp://en.wikipedia.org/wiki/International_Reactor_Innovative_and_Securehttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttp://www.vaec.gov.vn/userfiles/file/Fast%20Neutron%20Reactors.pdfhttp://www.vaec.gov.vn/userfiles/file/Fast%20Neutron%20Reactors.pdfhttp://www.vaec.gov.vn/userfiles/file/Fast%20Neutron%20Reactors.pdfhttp://www.vaec.gov.vn/userfiles/file/Fast%20Neutron%20Reactors.pdfhttp://www.vaec.gov.vn/userfiles/file/Fast%20Neutron%20Reactors.pdfhttp://www.vaec.gov.vn/userfiles/file/Fast%20Neutron%20Reactors.pdfhttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttps://netfiles.uiuc.edu/mragheb/www/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Heavy%20Water%20Reactor.pdfhttp://en.wikipedia.org/wiki/International_Reactor_Innovative_and_Securehttp://www.atomstroyexport.ru/show/press/469http://www.ecology.at/files/pr529_1.pdfhttp://books.google.ca/books?id=aChOvkuxTMkC&pg=PA436&lpg=PA436&dq=karena+reactor&source=bl&ots=nyxaIXTSRK&sig=cCLLp7uha5ASEqG7fOWythsshcU&hl=en&ei=sI_ETOqhC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