Light Water Reactors (LWR) was, covered in previous chapter, and the remaining types of nuclear power reactors – Heavy Water Reactors (HWR), High Temperature Gas-Cooled Reactors (HTGR), and Fast Neutron Reactors (FNR) will be covered under this chapter.

No Description
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
2.1 Enhanced   CANDU-6(EC6);
2.2 Advanced CANDU Reactors (ACR);
2.3 Advanced   Heavy Water Reactors (AHWR);
3.1 HTR-PM;
3.2 Pebble Bed Modular Reactors (PBMR); and
3.3 Gas   Turbine – Modular Helium Reactors (G7-MHR).
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


Heavy water reactors use heavy water as a neutron moderator. Heavy water is deuterium oxide, D2O. Neutrons in a nuclear reactor that uses uranium must be slowed down so that they are more likely to split other atoms and get more neutrons released to split other atoms.

The HWR concept allow the use of natural uranium as a fuel without the need for its enrichment, offering a degree of energy independence, especially if uranium is available for mining or for extraction as a byproduct of another industry such as gold mining or phosphate fertilizer production. However, it needs the installation of a heavy water D2O production capability which is a much simpler endeavor anyway since separating the light Isotopes (D from H) is much simpler than separating the heavy Isotopes.

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

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

The CANDU-9 (925-1300 MWe) was, developed from this also as a single-unit plant. It has flexible fuel requirements ranging from natural uranium through slightly enriched uranium, 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 plutonium or actinides separated from reprocessed PWR/BWR waste. A two year licensing review of the CANDU-9 design was successfully completed early in 1997, but the design has been shelved.

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 and Chinese 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 fuel options, plus 4.5 year construction and 60-year plant life (with mid-life pressure tube replacement). This is under consideration for new build in Ontario. AECL claims it as a Generation III design.

The Advanced CANDU Reactor (ACR), a 3rd generation reactor and is a more innovative concept. While retaining the low-pressure heavy water moderator, it incorporates some features 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 temperature and coolant pressure it has higher thermal efficiency.


2.2      Advanced CANDU Reactor (ACR):

The ACR-700 design was 700 MWe but is physically much smaller, simpler and more efficient as well as 40 percent cheaper than the CANDU-6. However, the ACR-1000 of 1080-1200 MWe (3200 MWt) is now the focus of attention by AECL.


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 volumes accordingly. 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 independent and fast shutdown systems. Units will be assembled from prefabricated modules cutting construction time to 3.5 years. ACR units can be built singly but are optimal in pairs. They will 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-licensing approval) but then withdrew after the first stage. In the USA, NRC lists the ACR-700 as being at pre application review stage. The first ACR-1000 unit could be operating in 2016 in Ontario.

The CANDU X or SCWR is a variant of the ACR, but with supercritical light water coolant (eg 25 MPa and 625ºC which represents 1157 Degree Fahrenheit) to provide 40 percent thermal efficiency. The size range envisaged is 350 to 1150 (662 to 2102 Degree Forhenheit) MWe depending on the number of fuel channels used. Commercialization envisaged after 2020.

2.3     Advanced Heavy Water Reactor (AHWR):

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


The calandria has about 450 vertical pressure tubes and the coolant is boiling light water circulated by convection. A large heat sink – “Gravity-driven water pool” – with 7000 cubic metres of water is near the top of the reactor building. Each fuel assembly has 30 Th-U-233 oxide pins and 24 Pu-Th oxide pins around a central rod with a burnable absorber. Burn-up of 24 GWd/t is, envisaged. It is designed to be self-sustaining in relation 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 is expected to utilize 65 percent of the energy of the fuel with two thirds of that energy coming from thorium via U-233.

Once it is operational each AHWR fuel assembly will have the fuel pins arranged in three concentric 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 at equilibrium discharge burn-up level. As well as U-233 some U-232 is formed and the highly gamma-active daughter products of this confer a substantial proliferation resistance.

In 2009, an export version of this design was announced: the AHWR-LEU. This will use low-enriched uranium plus thorium as fuel dispensing with plutonium input. About 39 percent 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 percent average fissile content of the U-Th fuel. While designed for closed fuel cycle this is not required. 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 inherent proliferation resistance. The AEC says, “The reactor is manageable with modest industrial infrastructure within the reach of developing countries.” In the AHWR-LEU the fuel assemblies 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.


High Temperature Gas cooled Reactors (HTGR) distinguished from other gas-cooled reactors by the higher temperatures attained within the reactor. Such higher temperatures might permit the reactor to be used as an industrial heat source in addition to generating electricity.

Among the future uses for which HTGR are being considered is the commercial generation of hydrogen from water. In some cases HTGR turbines run directly by the gas that is used as a coolant. In other cases steam or alternative hot gases such as nitrogen are produced in a heat exchanger to run the power generators. Recent proposals have favoured helium as the gas used as an HTGR coolant.

These reactors use helium as a coolant at up to 950ºC (1742 Degree Farhenheit) which either makes steam conventionally or directly drives a gas turbine for electricity and a compressor to return the 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 17 percent U-235. Layers of carbon and silicon carbide giving containment for fission products which is stable to 1600°C or more surround this. These particles may be arranged: in blocks as hexagonal ‘prisms’ of graphite, or in billiard ball-sized pebbles of graphite 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 China’s HTR-PM being built at Shidaowan in Shandong province. It has been developed by Tsinghua University’s INET which is the R&D leader and Chinergy Co., with China Huaneng Group leading the demonstration plant project.


This reactor will have two reactor modules, each of 250 MWt/ 105 MWe; using 9 percent enriched fuel (520,000 elements) giving 80 GWd/t discharge burnup. With an outlet temperature of 750 º C (1382 Degree Farhenheit), the pair will drive a single steam cycle turbine at about 40 percent thermal efficiency. This 210 MWe Shidaowan demonstration plant is to pave the way for an 18-unit (3x6x210MWe) full-scale power plant on the same site, also using the steam cycle. The plant life is envisaged as 60 years with 85 percent load factor.

3.2     Pebble Bed Modular Reactor (PBMR):

South Africa’s PBMR was being developed by a consortium led by the utility Eskom with Mitsubishi Heavy Industries from 2010.


It draws on German expertise. It aims for a step change in safety, economics and proliferation resistance. Production units would be 165 MWe. The PBMR will ultimately have a direct-cycle (Brayton cycle) gas turbine generator and thermal efficiency about 41 percent the helium coolant leaving the bottom of the core at about 900°C (1652 Degree Farhenheit) I and driving a turbine. Power is adjusted by changing the pressure in the system. The helium is passed through a water-cooled pre-cooler and intercooler before being returned to the reactor vessel. (In the Demonstration Plant, it will transfer heat in a steam generator rather than driving a turbine directly).

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

Performance includes great flexibility in loads (40-100 percent) with rapid change in power settings. Power density in the core is about one tenth of that in a light water reactor and if coolant circulation ceases the fuel will survive initial high temperatures while the reactor shuts itself down – giving inherent safety. Overnight capital cost (when in clusters of eight units) are 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):

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


The cylindrical core consists of 102 hexagonal fuel element columns of graphite blocks with channels for helium and control rods. Graphite reflector blocks are both inside and around the core. Half the core is, replaced every 18 months. Burn-up is about 100,000 MWd/t. It is being developed by General Atomics in partnership with Russia’s OKBM Afrikantov supported by Fuji (Japan). Initially it was to be used to burn pure ex-weapons plutonium at Seversk (Tomsk) in Russia. The preliminary design stage was completed in 2001, but the program has stalled since. Areva’s Antares is based on the GT-MHR.


Fast neutron reactors are a technological step beyond conventional power reactors. They offer the prospect of vastly more efficient use of uranium resources and the ability to burn actinides 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 isotope used in most reactors. If they are designed to produce more plutonium than they consume they are 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 Neutron Reactors and the IAEA’s INPRO program involving 22 countries (see later section) has fast 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 fast neutron 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 the world’s uranium resources – it could do this by a factor of about 60.  When those resources were perceived to be scarce several countries embarked upon extensive FBR development programs. However, significant technical and materials problems were encountered and geological exploration showed by the 1970s that uranium scarcity was not going to be a concern for some time. Due to both factors by the 1980s it was clear that FNRs would not be commercially competitive with existing light water reactors for some time. Here are some Fast Neutron Reactors:

4.1     Fast Breeder Reactor (FBR):

Several countries have research and development programs for improved Fast Breeder Reactors (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, and some have supplied electricity commercially. About 300 reactor-years of operating experience have been, accumulated.


Natural uranium contains about 0.7 percent U-235 and 99.3 percent U-238. In any reactor the 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 produce heat. In a fast neutron reactor this process is optimized so that it can ‘breed’ fuel, often using a depleted uranium blanket around the core. FBR can utilize uranium at least 60 times more efficiently than a normal reactor. They are however expensive to build and could only be justified economically if uranium prices were to rise to pre-1980 values well above the current market price. For this reason research work almost ceased for some years and that on the 1450 MWe European FBR has apparently lapsed. Closure of the 1250 MWe French Superphenix FBR after very little operation over 13 years also set back developments.

Research continues in India at the Indira Gandhi Centre for Atomic Research a 40 MWt fast breeder test reactor has been operating since 1985. In addition, the tiny Kamini there is employed to explore the use of thorium as nuclear fuel by breeding fissile U-233. In 2004 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 (the reactor-grade Pu being from its existing PHWR) and with a thorium blanket to breed fissile U-233. This will take India’s ambitious thorium program to stage two and set the scene for eventual full utilization of the country’s abundant thorium to fuel reactors.

Japan plans to develop FBR and its Joyo experimental reactor which has been operating since 1977 is now being boosted to 140 MWt. The 280 MWe Monju prototypes commercial FBR was connected to the grid in 1995, but was then shut down due to a sodium 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 Japan Standard Fast Reactor (JSFR) concept though with breeding ratio less than 1:1. This is a large 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 Mitsubishi FBR Systems (MFBR).

4.3     BN-600:

The Russian BN-600 fast breeder reactor at Beloyarsk has been supplying electricity to the grid since 1981 and has the best operating and production record of all Russia’s nuclear power units.


It uses uranium oxide fuel and the sodium coolant delivers 550°C at little more than atmospheric 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-600 to 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 is being built at Beloyarsk. It has considerable fuel flexibility – U+Pu nitride, MOX, or metal and with breeding ratio up to 1.3. It has much enhanced safety and improved economy – operating cost is expected to be only 15 percent more than VVER. It is capable of burning 2 tonnes of plutonium per year from dismantled weapons and will test the recycling of minor actinides in the fuel. The BN-800 has been sold to China and two units are due to start construction there in 2012.


However, the Beloyarsk-4 BN-800 is likely to be the last such reactor built (outside India’s thorium program), with a fertile blanket of depleted uranium around the core. Further fast reactors will have an integrated core to minimize the potential for weapons proliferation 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.


Pb-208 (54 percent of naturally occurring lead) is transparent to neutrons. A significant new Russian design from NIKIET is the BREST fast neutron reactor, of 300 MWe or more with lead as the primary coolant at 540°C, and supercritical steam generators. It is inherently safe and uses a high-density U+Pu nitride fuel with no requirement for high enrichment levels. No weapons-grade plutonium can be produced (since there is no uranium 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 and radiologically ‘hot’. Subsequently, any surplus plutonium which is not in pure form can be, used as the cores of new reactors. Used fuel can be recycled indefinitely with on-site reprocessing and associated facilities. A pilot unit is planned for Beloyarsk by 2020 and 1200 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 Nuclear from Italy and financed by Erratum. ELSY is a flexible fast neutron reactor which can use depleted uranium or thorium fuel matrices and burn actinides from LWR fuel. Liquid metal (Pb or Pb-Bi eutectic) cooling is at low pressure. The design was nearly complete in 2008 and a small-scale demonstration facility is planned. It runs on MOX fuel at 480°C and the molten lead is pumped to eight steam generators though decay heat removal is passive by convection.

4.7     PRISM:

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

Today’s PRISM is a GE-Hitachi design for compact modular pool-type reactors with passive cooling for decay heat removal. After 30 years of development it represents GEH’s Generation IV solution to closing the fuel cycle in the USA. Each PRISM Power Block consists of two modules of 311 MWe each, operating at high temperature – over 500°C.


The pool-type modules below ground level contain the complete primary system with sodium coolant. The Pu & DU fuel is metal and obtained from used light water reactor fuel. However, all transuranic elements are, removed together in the electrometallurgical reprocessing so that fresh fuel has minor actinides with the plutonium. Fuel stays in the reactor about six years, with one third removed every two years. Used PRISM fuel is recycled after removal of fission products. The commercial-scale plant concept part of an Advanced Recycling Centre uses three power blocks (six reactor modules) to provide 1866 MWe.

4.8     KALIMER (KAERI):

Korea’s KALIMER (Korea Advanced Liquid Metal Reactor) is a 600 MWe pool type sodium-cooled fast reactor designed to operate at over 500ºC. It has evolved from a 150 MWe version. It has a transmuter core and no breeding blanket is involved. Future development of KALIMER as a Generation IV type is envisaged.

Advanced water-cooled-reactor nuclear energy system concepts have been identified as part 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, the UK and the U.S. This activity resulted in the proposal of over thirty-eight specific reactor designs. The leading reactor designs can be, categorized into two general groups:

  • Near-Term Advanced Boiling Water and Pressurized Water Reactors with Passive-Safety; and
  • Longer-Term Advanced Water Reactors – e.g., Supercritical Water Reactor.

The first grouping of advanced Boiling Water Reactor (BWR) and Pressurized Water Reactors (PWR) systems can be represented by the Experimental Simplified Boiling Water Reactor (ESBWR)) and the Advanced Pressurized Water Reactor (AP1000), while the Supercritical Water Reactor (SCWR) is a unique example of the second grouping.

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

  • Advanced Gas-Cooled Reactors for High Temperatures (PBMR, MHGR, VHTR, GFR); and
  • 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 High Temperature Gas Reactor (VHTR) either with graphite pebbles or with prismatic graphite blocks as moderators.

The second grouping can be represented by the integral sodium-cooled fast reactor or the lead-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 very successful where they have been, built in Europe, Asia and the Pacific Rim, further evolution is needed to make new nuclear energy systems a more attractive option for deployment around the world. In particular, the next generation of nuclear energy systems must be able to be licensed, constructed, and operated in a manner that will provide a competitively priced supply of energy while satisfactorily addressing plant reliability, nuclear safety, waste disposal, proliferation resistance, and public perception concerns of the countries in which they are deployed.


  1. World Nuclear Association – Advanced Nuclear Power Reactors;
  2. Advanced Nuclear Energy Systems:  Heat Transfer Issues and Trends;
  1. GE Hitachi Nuclear Energy;
  2. Wikipedia:  Economic Simplified Boing Water Reactor;
  3. ATMEA: The ATMEA1      Reactor;
  4. Karena Reactor: Multiphase      Flow Dynamics 4 – Review;
  5. AES-92 for Belene: The      Mystery Reactor;
  6. MIR-1200;
  7. Wikipedia: International      Reactor Innovative and Secure;
  8. Heavy Water Reactors;      and
  9. Fast Neutron Reactors.

This chapter was published on “Inuitech – Intuitech Technologies for Sustainability” on May 20, 2011; and

This chapter was updated on 10 June 2020

Chapter 08