Very High Temperature Reactor (VHTR) Description
~ a next step in the evolutionary development of high-temperature gas-cooled reactors
~ can produce hydrogen from only heat and water by using thermochemical iodine-sulfur (I-S) process or
~ from heat, water, and natural gas by applying the steam reformer technology to core outlet temperatures greater than about 1000°C
~ a reference VHTR system that produces hydrogen is shown in the document
~ a 600 MWth VHTR dedicated to hydrogen production can yield over 2 million normal cubic meters per day
~ VHTR can also generate electricity with high efficiency, over 50% at 1000°C, compared with 47% at 850°C in the GTMHR or PBMR
~ co-generation of heat and power makes the VHTR an attractive heat source for large industrial complexes
~ VHTR can be deployed in refineries and petrochemical industries to substitute large amounts of process heat at different temperatures,
~ including hydrogen generation for upgrading heavy and sour crude oil
~ core outlet temperatures higher than 1000°C would enable nuclear heat application to such processes as
~ steel, aluminum oxide, and aluminum production

~ VHTR is a graphite-moderated, helium-cooled reactor with thermal neutron spectrum
~ it can supply nuclear heat with core-outlet temperatures of 1000°C
~ the reactor core type of the VHTR can be a prismatic block core such as the operating Japanese HTTR,
~ or a pebble-bed core such as the Chinese HTR-10
~ for electricity generation, the helium gas turbine system can be directly set in the primary coolant loop, which is called a direct cycle
~ for nuclear heat applications such as process heat for refineries, petrochemistry, metallurgy,
~ and hydrogen production, the heat application process is generally coupled with the reactor through an
~ intermediate heat exchanger (IHX), which is called an indirect cycle

Technology Base for the VHTR
~ evolves from HTGR experience and extensive international databases that can support its development
~ the basic technology for the VHTR has been well established in former HTGR plants, such as Dragon,
~ Peach Bottom, AVR, THTR, and Fort St Vrain and is being advanced in concepts such as the GT-MHR and PBMR
~ the ongoing 30-MWth HTTR project in Japan is intended to demonstrate the feasibility of reaching outlet temperatures up to 950°C coupled
~ to a heat utilization process
~ the HTR-10 in China will demonstrate electricity and co-generation at a power level of 10 MWth
~ the former projects in Germany and Japan provide data relevant to VHTR development
~ steam reforming is the current hydrogen production technology
~ the coupling of this technology will be demonstrated in large scale in the HTTR program but still needs complementary R&D for market introduction
~ R&D on thermochemical I-S process is presently proceeding in the laboratory-scale stage

Technology Gaps for the VHTR

Design parameters for the VHTR:
~ Reactor power: 600 MWth
~ Coolant inlet/outlet temperature: 640/1000°C
~ Core inlet/outlet pressure: Dependent on process
~ Helium mass flow rate: 320 kg/s
~ Average power density: 6–10 MWth/m3
~ Reference fuel compound: ZrC-coated particles in blocks, pins or pebbles
~ Net plant efficiency: >50%

Challenges
~ demonstrating the viability of the VHTR core requires meeting a number of significant technical challenges
~ novel fuels and materials must be developed that:
~ permit increasing the core-outlet temperatures from 850°C to 1000°C and preferably even higher
~ permit the maximum fuel temperature reached following accidents to reach 1800°C
~ permit maximum fuel burnup of 150–200 GWD/MTHM
~ avoid power peaking and temperature gradients in the core, as well as hot streaks in the coolant gas

Technology Gaps for the VHTR
~ process-specific R&D gaps exist to adapt the chemical process and the nuclear heat source to each other with regard to
~ temperatures, power levels, and operational pressures
~ heating of chemical reactors by helium is different from current industrial practice and needs specific R&D and demonstration
~ qualification of hightemperature alloys and coatings for resistance to corrosive gases like
~ hydrogen, carbon monoxide, and methane will be needed
~ the viability of producing hydrogen using the iodinesulfur (I-S) process still requires pilot- and
~ large-scale demonstration of the three basic chemical reactions and development of corrosion-resistant materials
~ any contamination of the product will have to be avoided
~ development of heat exchangers, coolant gas ducts, and valves will be necessary for isolation of the nuclear island from the production facilities
~ this is especially the case for isotopes like tritium, which can easily permeate metallic barriers at high temperatures
~ performance issues for the VHTR include development of a high-performance helium turbine for efficient generation of electricity
~ modularization of the reactor and heat utilization systems is another challenge for commercial deployment of the VHTR

Qualification of TRISO Fuel
~ the increase of the helium core-outlet temperature of the VHTR results in an increase of the fuel temperature and reduced margins in case of core heatup accidents
~ fuel particles coated with silicon-carbide are used in HTGRs at fuel temperatures of about 1200°C
~ irradiation testing is required to demonstrate that TRISO-coated particles can perform acceptably
~ at the high burnup and temperature associated with the VHTR
~ following irradiation, hightemperature heating (safety) tests are needed to determine that there is no degradation
~ in fuel performance under accident heatup conditions up to 1600°C as a result of the more demanding irradiation service conditions
~ these fuel demonstration activities would require about 5 to 7 years to complete following fabrication of samples
~ complete fuel qualification would require an additional 5 to 7 years in which statistically significant
~ production scale fuel is irradiated to confirm the performance of the fuel from the production facility
~ irradiation facilities and safety test facility exist worldwide,
~ and an integrated coordinated fuel development program could shorten development times by one-third

ZrC Coatings for TRISO Fuel
~ qbove a fuel temperature of 1200°C, new coating materials such as zirconium-carbide and/or improved coating techniques should be considered
~ use of ZrC in HTGRs enables an increase in power density and an increase in power size under the same coolant outlet temperature
~ and allows for greater resistance against chemical attack by the fission product palladium
~ the limited fabrication and performance data on ZrC indicates that although it is more difficult to fabricate,
~ it could allow for substantially increased operating and safety envelopes (possibly approaching 1800°C)
~ only laboratory-scale fabrication of ZrC-coated particle fuel has been performed to date
~ research into more economical commercial-scale fabrication routes for ZrC-coated particle fuels,
~ including process development at production scale, is required
~ advanced coating techniques or advanced processing techniques (automation) should be considered
~ prrocess development on production-scale coating is required
~ irradiation testing and high-temperature heating (safety) tests are needed to define operation and safety envelopes/
~ limits for this fuel, with the goal of high burnup (>10% FIMA and high-temperature (1300–1400°C) operation
~ the facilities used for TRISO-coated particle testing can also be used for ZrC-coated fuel development
~ these activities would require 10 to 15 years to complete and could be performed at facilities adapted
~ from those available around the world currently used for SiC-based coated particle fuel

Burnable Absorbers
~ increasing the allowable fuel burnup requires development of burnable absorbers for reactivity control
~ the behavior of burnable absorbers needs to be established (e.g., irradiation dimensional stability, swelling, lifetime) under the design service conditions of the VHTR

Carbon-Carbon Composite Components
~ development of carbon-carbon composites is needed for control rod sheaths,
~ especially for the VHTR based on a prismatic block core, so that the control rods can be inserted to the high-temperature areas entirely down to the core
~ promising ceramics such as fiber-reinforced ceramics, sintered alpha silicon-carbide, oxide-composite ceramics,
~ and other compound materials are also being developed for other industrial applications needing highstrength, high-temperature materials
~ planned R&D includes testing of mechanical and thermal properties, fracture behavior, and oxidation;
~ post irradiation heat-up tests; and development of models of material behavior and stress analysis code cases considering anisotropy
~ the feasibility of using superplastic ceramics in VHTR components will be investigated by studying the effects
~ of neutron irradiation on superplastic deformation mechanisms
~ testing of core internals is envisioned to take 5 to 10 years at any of the test reactors worldwide

Pressure Vessel Materials
~ to realize the goal of core outlet temperatures higher than 1000°C,
~ new metallic alloys for reactor pressure vessels have to be developed
~ at these core-outlet temperatures, the reactor pressure vessel temperature will exceed 450°C
~ LWR pressure vessels were developed for 300°C service, and the HTTR vessel for 400°C
~ hasteloy-XR metallic materials are used for intermediate heat exchanger and hightemperature
~ gas ducts in the HTTR at core-outlet temperatures up to about 950°C,
~ but further development of Ni-Cr-W super-alloys and other promising metallic alloys will be required for the VHTR
~ the irradiation behavior of these superalloys at the service conditions expected in the VHTR will need to be characterized
~ such work is expected to take 8 to 12 years and can be performed at facilities available worldwide
~ an alternate pressure vessel allowing for larger diameters and ease of transportation, construction, and dismantling would be the prestressed cast-iron vessel,
~ which can also prevent a sudden burst due to separation of mechanical strength and leak tightness
~ the vessel could also include a passive decay heat removal system with enhanced efficiency

Heat Utilization Systems Materials
~ internal core structures and cooling systems, such as intermediate heat exchanger, hot gas duct, process components, and
~ isolation valve that are in contact with the hot helium can use the current metallic materials up to about 1000°C core-outlet temperature
~ for core-outlet temperatures exceeding 1000°C, ceramic materials must be developed
~ piping and component insulation also requires design and materials development

VHTR Reactor Systems R&D /Core Internals
~ core internal structures containing the fuel elements such as pebbles or blocks are made of high-quality graphite
~ the performance of high-quality graphite for core internals has been demonstrated in gascooled pilot and demonstration plants,
~ but recent improvements in the manufacturing process of industrial graphite have shown improved oxidation resistance and better structural strength
~ irradiation tests are needed to qualify components using advanced graphite or composites to the fast fluence limits of the VHTR

VHTR Balance-of-Plant R&D
~ the VHTR balance-of-plant is determined by the specific application, which can be thermochemical processes, dedicated electricity production or cogeneration
~ all components have to be developed for temperatures well above the present state of the art and depend on a comprehensive material qualification activity
~ failure mechanisms such as creep, fretting, and ratcheting have to be studied in detail,
~ precluded with design, and demonstrated in component tests
~ specific components such as IHX, isolation valves, hot gas ducts with low heat loss, steam reformers,
~ and process-related heat exchangers have to be developed for use in the modular VHTR, which mainly uses only one loop
~ this leads to much larger components than formerly developed and a new design approach by modularization of the component itself
~ low pressures are necessary or preferable for many processes
~ alternate coolants for the intermediate loop such as molten salt should be adapted where needed
~ process-specific components will need to be tested
~ other applications will require different components such as helium-heated steam crackers, distiller columns, and superheaters

I-S Process Subsystem
~ the development and qualification of an I-S process subsystem is needed
~ this is discussed in the Crosscutting Energy Products R&D section

Analysis Methods
~ extension and validation of existing engineering and safety analysis methods is required to include
~ new materials, operating regimes, and component configurations in the models
~ new models need to be developed for the VHTR with balance of plant consisting of thermochemical process and other energy applications

VHTR Safety R&D
~ passive heat removal systems should be developed to facilitate operation of the VHTR,
~ with a final goal of simple operation and transparent safety concepts
~ demonstration tests should be performed on the VHTR to verify the system’s passive characteristics,
~ which have a lower margin between operational temperatures and the limits for fuel and materials
~ analysis and demonstration of the inherent safety features of the VHTR are needed,
~ and could potentially draw on development and demonstration of earlier INTD gas reactors
~ additional safety analysis is necessary with regard to nuclear process heat applications in an industrial environment
~ the safe isolation of the reactor system after failures in the heat delivery system is an essential issue for demonstration of IHX
~ and hot gas valve tightness after depressurization of the secondary circuit
~ full-scale tests of valves and IHX modules will be necessary
~ design basis and severe accident analyses for the VHTR will need to include phenomena such as chemical attack of graphitic core materials,
~ typically either by air or water ingress
~ adequacy of existing models will need to be assessed, and new models, may need to be developed and validated

VHTR Fuel Cycle R&D

Disposal of Once-Through Fuel and Graphite
~ VHTR assumes a once-through, LEU (<20% 235U) fuel cycle
~ like LWR spent fuel, VHTR spent fuel could be disposed of in a geologic repository or conditioned for optimum waste disposal
~ the current HTGR particle fuel coatings form an encapsulation for the spent fuel fission products
~ that is extremely resistant to leaching in a final repository
~ however, as removed from the reactor, the fuel includes large quantities of graphite, and
~ research is required to define the optimum packaging form of spent VHTR fuels for long-term disposal
~ radiation damage will require graphite replacement every 4 to 10 years
~ an optimized approach for dealing with the graphite (i.e., recycle, low-level waste, remain integral with spent fuel) remains to be defined

Fuel Recycling
~ recycling of LWR and VHTR spent fuel in a symbiotic fuel cycle can achieve significant reductions in waste quantities and radiotoxicity
~ because of the VHTR’s ability to accommodate a wide variety of mixtures of fissile and fertile materials without significant modification of the core design
~ this flexibility was demonstrated in the AVR test reactor in Germany and is a result of the ability of gas reactors to decouple
~ the optimization of the core cooling geometry from the neutronics
~ for an actinide burning alternative, specific Pu-based driver fuel and transmutation fuel containing minor actinides would have to be developed
~ this fuel can benefit from the above mentioned R&D on SiC and ZrC coating but will need more R&D than LEU fuel

Desalination and District Heating Interface R&D
~ this area of R&D considers desalination to produce fresh water
~ with regard to desalination, multiple approaches are possible either through direct use of low temperatures heat (120°C) or
~ through optimized reverse osmosis processes
~ with regard to district heating, a nuclear-supplied district heating network has operated for almost two decades in Switzerland
~ this provides a valuable benchmark for evaluating district heating applications
~ many cities in Eastern Europe, Russia, and the Former Soviet Union are already equipped with a district heating infrastructure
~ in the Brayton cycle, coolant temperatures in the heat exchanger range from 150°C down to 30°C and discharge heat to the low-temperature heat sink
~ in thermochemical processes such as the I-S process, heat in the range of 100–150°C is available
~ thus, the Brayton cycle and thermochemical processes for hydrogen production may potentially be combined with desalination,
~ district heating, or numerous other process-heat applications as a co-generation system without reducing the thermal efficiency of electricity generation or hydrogen production
~ R&D is recommended to explore the impact on the overall plant design and optimization

»Cogeneration Technologies
~ Cooler, Cleaner, Greener Power & Energy Solutions™ project development services
~ located in Houston, provides project development services that generate clean energy and significantly reduce greenhouse gas emissions and carbon dioxide emissions
~ DOE/the Generation IV initiative in 2000, then the Generation-IV International Forum (GIF)
~ Argentina,Brazil,Canada,France,Japan,Korea,South Africa,Switzerland,UK and the United States

Six nuclear systems were selected:
~ Gas-Cooled Fast Reactor
~ Lead-cooled Fast Reactor
~ Molten Salt Reactor
~ Sodium-Cooled Fast Reactor
~ Supercritical-Water-Cooled Reactor
~ Very High Temperature Reactor


»Very High Temperature Reactor (VHTR) by Idaho National Laboratory
~ a graphite-moderated, helium-cooled reactor with a once-through uranium fuel cycle
~ supplies heat with core outlet temperatures of 1,000 degrees Celsius,
~ which enables applications such as hydrogen production or process heat for the petrochemical industry or others
~ the reference reactor is a 600 MWth core connected to an intermediate heat exchanger to deliver process heat
~ core can be a prismatic block core such as the operating Japanese HTTR, or
~ a pebble-bed core such as the operating Chinese HTR-10
~ for hydrogen production, the system supplies heat that could be used efficiently by the thermochemical iodine-sulfur process
~ system is designed to be a high-efficiency system that can supply process heat to a broad spectrum of high-temperature and energy-intensive, nonelectric processes
~ system may incorporate electricity generating equipment to meet cogeneration needs
~ system also has the flexibility to adopt uranium/plutonium fuel cycles and offer enhanced waste minimization
~ the VHTR offers a broad range of process heat applications and an option for high-efficiency electricity production,
~ while retaining the desirable safety characteristics offered by modular high-temperature gas-cooled reactors

VHTR Reactor Pursued in U.S. With Priority
~ Idaho National Laboratory (INL) to issue a contract to Westinghouse Electric Company for the pre-conceptual design of the NGNP,
~ to AREVA NP and General Atomics
~ they perform complimentary engineering studies in the areas of technology and design tradeoffs, initial cost estimates and selected plant arrangements

»High and very high temperature reactors (HTR/VHTR)

»Generation IV Technology / VHTR
~ designed to be a high-efficiency system, which can supply electricity and process heat to a broad spectrum of high-temperature and energy-intensive processes
~ the reference reactor is a 600 MWth core connected to an intermediate heat exchanger to deliver process heat
~ the reactor core can be a prismatic block core or a pebble-bed core according to the fuel particles assembly
~ fuel particles are coated with successive material layers, high temperature resistant, then formed either into fuel compacts embedded in graphite block for the prismatic block-type core reactor,
~ or formed into graphite coated pebbles
~ the reactor supplies heat with core outlet temperatures up to 1,000 degrees Celsius,
~ which enables such applications as hydrogen production or process heat for the petrochemical industry
~ as a nuclear heat application, hydrogen can be efficiently produced from only heat and water by using thermochemical iodine-sulfur process,
~ or high temperature electrolysis process or with additional natural gas by applying the steam reformer technology
~ offers a high-efficiency electricity production and a broad range of process heat applications,
~ while retaining the desirable safety characteristics in normal as well as off-normal events
~ solutions to adequate waste management will be developed
~ the basic technology for the VHTR has been well established in former High Temperature Gas Reactors plants,
~ such as the US Fort Saint Vrain and Peach Bottom prototypes, and the German AVR and THTR prototypes
~ the technology is being advanced through near or medium term projects lead by several plant vendors and national laboratories,
~ such as: PBMR, GT-HTR300C, ANTARES, NHDD, GT-MHR and NGNP in South Africa, Japan, France, South Korea and the United States
~ experimental reactors: HTTR (Japan, 30 MWth) and HTR-10 (China, 10 MWth) support the advanced concept development,
~ and the cogeneration of electricity and nuclear heat application

VHTR / THE FRENCH ATOMIC ENERGY COMMISSION (CEA) R&D PROGRAM
~ the material technologies investigated are those of respectively,
~ medium temperature (~ 450-650°C) for vessel structures and cold internals,
~ high temperature (~ 650-950°C) for primary circuit, recuperator, intermediate heat exchanger, turbine component and
~ very high temperature (~ 1000-1650°C) for core and hot gas facing components
Graphite and Structural Composites
~ In the VHTR, the core structural elements are made from graphite, in order to provide
~ neutron moderation and high temperature structural support to the fuel
~ the graphite components of the reactor are the permanent inside and outside reflectors,
~ the core blocks (in the case of a block reactor), and the core supports
~ ASME Code Section III, Division 2, Subsection CE for graphite core support structures must be updated and approved for the VHTR temperatures
~ graphites which can be used in VHTR should be reasonably dense (1.75/1.80 g.cm-3),
~ should be well graphitized (graphitization temperature > 2700°C), as indicated by a thermal conductivity higher than 100 W.m-1.K-1 at room temperature (RT)
~ they should have a low neutron cross section (˜ 4 mbarns); and
~ the total neutron absorbing impurities should be lower than 3 ppm boron equivalent
~ the impurities which could lead to operational problems (Li, B, Co), and high decommissioning costs (N, Cl, Ni) should be reduce to a minimum
~ they should have high dimensional stability under irradiation,
~ indicated by a coefficient of thermal expansion (20°C – 120°C) between 4 – 5.5 10-6.K-1;
~ this implies to use relatively isotropic graphite
~ the irradiation time, over the irradiation temperature of interest, for the graphite to return to its original volume should be as long as possible
~ the graphite should have a tensile strength of about 20 MPa
~ the air reactivity should be measured to ensure that the rate is not higher than 0.2 mg.g-1.h-1
~ in the framework of the HTR-materials european programme (HTR-M), different graphite grades have been chosen to be studied
~ three of these grades have been manufactured by UCAR, three other grades have been manufactured by SGL
~ two isomolded graphites supplied by Toyo Tanso will be studied
~ all these graphites are irradiated in the High Flux Reactor (HFR) of Petten in the temperature range of 700-750°C
~ with an objective of a damage level of 8 dpa.graphite (1.06 1022 cm-2 [E > 0.1 MeV])
~ the studied properties after irradiation will be: dimensional changes; grain morphology;
~ coefficient of thermal expansion (CTE), heat capacity and thermal conductivity up to the irradiation temperature;
~ Young modulus, compressive and flexural strength, fracture toughness at room temperature and at the irradiation temperature
~ generally the most important property change for the core lifetime integrity is dimensional changes
~ when it is irradiated, graphite shows initially a short period of swelling followed by a larger shrinkage with irradiation fluence
~ then this shrinkage slows down and there is a point of “turn around” before the graphite swells back to its original volume and beyond
~ the graphite dimensional changes depend on the irradiation temperature, the coke anisotropy, the grain size and the graphitization temperature
~ dimensional changes are lower for the more isotropic coke and high crystallinity graphites;
~ this implies large crystallite size and high graphitization temperature (2800-3000°C)
~ another property which strongly changes with neutron irradiation is the thermal conductivity
~ neutron induced degradation of the thermal conductivity occurs at doses as low as 10-3 dpa.g
~ the degradation of thermal conductivity decreases with increasing irradiation temperature
~ the Young modulus increases quickly at the beginning of the irradiation;
~ then a saturation of this phenomenon appears at a neutron fluence between 0.5 and 2.1021 cm-2
~ on the other hand, carbon fiber re-enforced carbon (CFRC) composites with allowable temperatures up to 1800 °C and
~ other ceramic composite materials (SiC/graphite and SiC/SiC) have been proposed for the several subcomponents in the control rod assembly
~ the good performances of carbon/carbon composites are based on their high strength and their high thermal conductivity
~ CEA has launched a R&D programme to study the neutron irradiation effects on carbon-carbon composites, especially thermal conductivity and dimensional changes
»Graphite

»Very High Temperature Reactor
~ there are early signs that the nature of the nuclear reactor market might be changing
~ Finland is building based on Framatome ANP's EPR design for a "fifth nuclear reactor"
~ France also plans to build an EPR
~ Bulgaria has discussed building a new nuclear reactor at Belene
~ the United States is funding a program called Nuclear Power 2010 that seeks to build at least two nuclear power reactors by the mid 2010's
~ supporting this has been proposed Federal energy legislation
~ meeting the target would be a challenging task and the proposed legislation is still being debated

»VHTR and The Principles of Nuclear Power
The Principles of Nuclear Power
~ in naturally occurring uranium, 0.7% of uranium is of a particular type (isotope) of uranium (U235)
~ which spontaneously splits (fissile material) to emit a tiny particle (a neutron)
~ if this neutron hits another U235 atom, it too will split (a fission) to produce two more neutrons (chain reaction)
~ if the concentration of U235 is sufficient (a critical mass), the process will be self-sustaining (the plant is `critical'),
~ producing large quantities of heat in the `core' of the reactor

~ two important ingredients are needed to control the process and to utilise the heat, the moderator and the coolant
~ a moderator is a substance which neutrons collide with but `bounce off' without absorbing too much energy and without itself being split
~ it controls the amount of neutrons escaping from the core before they have hit another U235 atom
~ a good moderator is one which absorbs the least energy and does not absorb the neutrons before they split another uranium atom
~ graphite is an excellent moderator; ordinary water is a poorer moderator but is much cheaper
~ if water is used, the U235 content must be increased (enrichment) to about 3 per cent to allow a chain reaction to take place
~ a rare isotope of hydrogen (deuterium) can be used to make so-called heavy water (deuterium is twice the weight of normal hydrogen) and this is also an excellent moderator

~ in so-called fast (breeder) reactors (as opposed to the thermal reactors described above),
~ no moderator is used and some of the neutrons escape the core and strike a `jacket' of uranium
~ where they convert the unused part of the uranium, U238, to fissile material, plutonium, which can be used as a reactor fuel
~ the jacket is processed to isolate the plutonium for use in more fast reactors
~ the attraction of this design is obvious, it can use almost 100 per cent of naturally occurring uranium instead of the 0.7 per cent thermal reactors achieve
~ the disadvantage is equally obvious: it requires the separation, transport and widespread use of the material used to make nearly all nuclear weapons
~ and is regarded as a serious proliferation risk
~ the technical attractions of the design have lead to huge amounts of public money being spent on this technology
~ however, in practice, all prototype plants have proved most unreliable and the technology is now all but abandoned

~ in order to produce electricity, the heat in the core has to be transferred to a fluid (a liquid or a gas), the coolant
~ the heat will expand the fluid (boil it if it is water) and the force of the expanding gas can be used to drive a turbine generator to produce electricity
~ this principle of transferring heat from a `boiler' to a turbine generator is the same for all types of thermal power station
~ whether it uses nuclear or fossil fuel
~ the coolant can go directly from the core to the turbine generator or there can be an intermediate stage
~ where the coolant goes through a heat exchanger to produce steam in a second circuit
~ liquids are much denser than gases and so a given volume of liquid can cool much more efficiently than the same volume of gas,
~ so if the coolant circuit with a liquid cooled reactor breaks, the plant will only be cooled by gases, that is, steam and air,
~ and the plant could over-heat catastrophically

~ ordinary water is a common, cheap coolant for power plants of all types, including nuclear power
~ its primary safety disadvantage in a nuclear power plant is that if it escapes,
~ the reactor will not be properly cooled (loss of coolant accident, or LOCA)
~ water can also be corrosive and will require expensive materials to prevent damage to the coolant pipes
~ however, water coolant requires much less volume of materials because of its greater efficiency in cooling than gas
~ so pressurized water reactors (PWRs) of the type built at Koeberg in South Africa, which use water as the coolant,
~ are much more compact than, for example, the British designs of gas-cooled reactor
~ of the gas coolants possible, carbon dioxide was used in the British power plant designs, but while this is cheap, it is somewhat corrosive
~ helium is entirely inert, but is expensive so leakage has to be avoided

~ of the many possible technologies, two are of particular relevance to South Africa,
~ the two existing civil nuclear power reactors at Koeberg and the PBMR
~ The Koeberg plants are each 900 MW (1 megawatt (MW) is 1 million kilowatts (kW))
~ they are known as pressurised water reactors (PWRs) because the coolant is maintained as liquid
~ despite being at about 300°C by keeping it at very high pressures
~ this coolant is passed through a heat exchanger in which the energy is transferred to a second circuit
~ in which water is boiled and drives the steam turbine generator

~ ordinary water is used as the moderator and as a result, uranium enriched to about 3 per cent is required
~ The PWR is the most widely used design of nuclear reactor in the world and
~ just under half the 430 nuclear power plants in the world are of this design
~ the main supplier is Westinghouse and its design has been adopted by Framatome (the Koeberg supplier), Siemens and Mitsubishi
~ the PWR is a direct descendant of submarine propulsion units and, as a result,
~ its operating schedule is planned around annual stoppages when the plant is refuelled and maintenance is carried out
~ typically, a quarter of the fuel rods are replaced each year,
~ because the concentration of U235 is no longer great enough to maintain full power operation

~ the PBMR uses helium as the coolant and graphite as the moderator and is one of a number of designs
~ that come under the general classification of High Temperature (Gas-Cooled) Reactors, HTGRs or HTRs
~ the use of helium and graphite gives it several intrinsic safety and technical advantages over, say, the PWR
~ as noted above, the use of a gaseous coolant reduces the risk from loss of coolant accidents
~ being inert, helium can be used at very high temperatures without concerns about corrosion

~ the use of a good moderator like graphite increases the efficiency with which the uranium is used
~ with HTRs, fuel is made in ceramic pellets (or pebbles) which can also withstand very high temperatures,
~ compared to a PWR where the fuel is in the form of rods of uranium oxide contained in a metal cladding
~ with HTRs, the moderator is in the form of a coating for the fuel and is an integral part of it,
~ unlike the PWR where the water flows past the fuel
~ this gives some safety advantages as the moderator which controls the reactor cannot be separated from the fuel
~ this combination of helium coolant, graphite moderator and ceramic fuel allows the reactor to operate at very high temperatures, 750ºC compared to 300ºC in a PWR. This in turn means that a much higher proportion of the energy from the core can be turned into electricity (the thermal efficiency), 40 per cent compared to 34 per cent for a PWR. It also means that a much higher proportion of the U235 can be split, giving high fuel `burn-up'. This means that the reactors are more economical in their use of uranium and create a much lower volume of used, or `spent' fuel.

~ all high temperature reactors built to date have used highly enriched uranium (HEU) - more than 90 per cent U235
~ while this may lead to good uranium utilization, such material is a serious weapons proliferation risk
~ South Africa's nuclear bombs were built using HEU
~ the use of such a material as a basis for nuclear power plants to be exported round the world would raise huge concern on proliferation grounds
~ and it is unlikely that the international community would allow South Africa to go ahead using such material
~ for its PBMR, Eskom plans to use 7-8 per cent enriched uranium, very different to the type of fuel used in HTRs so far

~ like most purpose-designed reactor types, but unlike the submarine-derived PWR,
~ the PBMR would avoid the need for an annual shut-down for re-fuelling, by re-fuelling while the plant is operating, on-line
~ in theory, this should mean that extra power can be produced
~ in practice, on-line refuelling has not always worked out well because the machines for doing it are complex,
~ expensive and prone to break-down
~ qlso, the time required for maintenance, which is carried out at the same time as refuelling,
~ usually exceeds the time required for re-fuelling so on-line refuelling would not reduce the amount of time the plant is off-line

~ For example, in Britain, the Advanced Gas-Cooled Reactor (AGR) was designed to refuel on-line, at full power
~ but more than 20 years after the first plant went into service, the regulatory authorities still do not allow refuelling
~ at full power because of safety concerns
~ ironically, in 1965 when the AGR was chosen, it was the extra output that was expected to be produced because of on-line refuelling,
~ that swung the economic case in favour of the AGR over US designs
~ tis reduced the overall generation cost of the AGR by a small fraction of a penny
~ this experience will not necessarily be repeated in South Africa but it does demonstrate that refuelling on-line can be a difficult process
~ and that any projected economic advantages to on-line refuelling should be treated with some remaining sceptical

Europe
~ 10 countries have built nuclear power plants
~ Austria closed its plant without operating it after a referendum
~ Italy closed its three plants after a referendum
~ Sweden is committed to closing its plant early after a referendum
~ the newly elected German government has committed itself to phasing out nuclear power
~ The Netherlands and Switzerland are also likely to phase out nuclear power
~ the Spanish government ordered the abandonment of work on several unfinished plants in the 1980s
~ as argued above, new nuclear orders in Britain are not feasible,
~ leaving only Finland and France as the only countries where new orders are possible, although not inevitable
~ France has spent huge amounts of money developing its own nuclear capability and
~ it is inconceivable that, if orders were placed, it would not use French companies

Turkey
~ has talked about placing nuclear orders and frequently, deals are said to have been imminent
~ it seems unlikely that Turkey will be a major market for nuclear power in the next decade

North America
~ no orders not subsequently cancelled have been placed since 1974
~ Canada has developed its own technology which is now running into severe problems on the economics and safety side with several units shut down for several years as a result
~ it is barely conceivable that any new orders would be placed
~ In the USA, more than 100 nuclear orders were cancelled, losing consumers billions of dollars
~ as in Canada, the electricity industry is being liberalised and many existing nuclear plants are being categorized as stranded assets
~ the two Mexican units took more than 20 years to build and cost over-runs were huge
~ given this poor record, new orders for nuclear power in any of these countries are not feasible

South America
~ Brazil and Argentina have built nuclear power plants
~ Argentina has two operating plants and has been struggling to finance completion of a third plant, of Canadian design for more than 20 years
~ Brazil has one operating nuclear plant which, over a 20 year life, has an average availability of about 20 per cent
~ it may complete a second plant of German design which started construction in 1975 and will cost about US$9bn,
~ making it about the most expensive nuclear plant built
~ these countries are unlikely to want to repeat their sad experience with nuclear power, nor are their neighbous likely to launch new programs

Africa
~ only South Africa is actively pursuing nuclear power and the chances of nuclear sales outside South Africa are minimal

Asia
~ this leaves only Asia as a possible market for nuclear power!
~ the two giants of the continent are India and China, both with nuclear power programmes
~ India and Pakistan both acquired nuclear power plants in the 1960s but after India exploded a nuclear bomb in 1975, all international nuclear contacts were cut
~ as a result it has tried to develop its own designs based on the plant it bought from Canada
~ it now has about 10 small (200 MW) plants in service
~ all have seriously overrun their construction time and cost forecasts and have been hopelessly unreliable
~ India is now trying to buy a plant from Russia, but it is unlikely that either side has the cash to carry out this project
~ Pakistan has recently bought a small plant from China of Chinese design
~ like India, its poor record on nuclear proliferation makes it largely impossible for Western countries to do business there with nuclear technology
~ China has, for the past 20 years, had ambitious plans to build nuclear power plants of imported design and of its own design
~ these have resulted in few orders so far: two plants are in service of French design,
~ two more French plants are on order and two Canadian plants are on order
~ one plant of Chinese design, a 300 MW PWR, is in service, but is currently off-line with serious equipment problems
~ one plant of this design was sold to Pakistan and China is planning to build further units of this basic design, but double the size
~ all nuclear vendors are active in China because of the potential size of the market,
~ but it is doubtful whether China can finance a significant nuclear power program
~ Japan has developed a number of its own nuclear technologies, but none of these has been ordered for commercial use
~ all its operating plants are of US design and Japanese companies such as Mitsubishi, Hitachi and Toshiba
~ have spent large sums of money over the past 30 years developing an understanding of these technologies
~ as well as manufacturing facilities for them
~ while Japan now has a large number of operating plants (53 at the beginning of 1999),
~ public opposition and problems in finding sites due to seismic activity mean that further orders are now very difficult
~ there is no room on established sites for further plants and, now, only two plants are under construction
~ if Japan does order further plants, they will almost certainly be more units of US design or units using a new Japanese design
~ South Korea and Taiwan have nuclear power plants in service
~ Korea has 14 plants in service and another 3 under construction
~ it has expended a large amount of effort transferring US technology in and has built up full manufacture facilities
~ it is highly unlikely that future nuclear orders would not be supplied using these facilities
~ Taiwan has six plants in service and two on order
~ when these two plants are complete, there will be little scope for further nuclear plants
~ other Asian countries, such as Thailand and Indonesia have, for 20 years or more,
~ discussed the possibility of ordering nuclear plants
~ however, there is little to suggest that these discussions will soon be turned into nuclear orders

Material Science /VHTR

»Nevada RD

»High-Temperature Nickel-Based Alloys for VHTR Applications
~ VHTR is a helium-cooled reactor operating with outlet temperatures exceeding 950°C
~ high yields for energy generation and the ability to supply high-temperature process heat for hydrogen production are the advantages
~ high temperatures impose challenging design requirements on structural materials, particularly for the intermediate heat exchangers
~ nickel-based alloys are the most suitable materials for high temperatures, particularly Inconel Alloy 617 and Haynes 230
~ however, these alloys are not fully ASME code-qualified for nuclear applications
~ basic data are needed to achieve a complete understanding of their behavior at high temperatures
~ researchers must investigate the mechanical properties of these alloys and study surface/subsurface corrosion effects caused by helium impurities
~ as component integrity must be demonstrated over the entire operational lifetime on the order of 100,000 hours,
~ the evolution of properties over time must be taken into account, specifically the effects of thermal aging


Hydrogen & VHTR

»The Role of Hydrogen in a Future Sustainable Energy System
~ starts from the assumption that if renewable energy is to play a significant role in the world economy
~ it will be necessary to develop effective ways of storing it as a fuel
~ so that it can be used when and where it is needed
~ considering hydrogen as a candidate storage medium fuel, whose suitability for a given situation would be determined by a sustainability analysis

Hydrogen pipelines
~ In Europe hydrogen distribution networks by pipelines have existed for several decades
~ the Rhein-Ruhr network with a total length of 240 km
~ the Benelux-France network of Air Liquide with 966 km
~ industrial hydrogen pipeline networks are already in operation (1500 km) at several locations in Europe,
~ in refinery complexes and industrial clusters such as the Ruhr area

Alternatives
~ alternatives in the introduction phase are distribution of liquid hydrogen by ship and truck, and on-site generation of hydrogen

VHTR /EU

»The Gas Cooled Reactors Technology Platform

Primary Technical, Economic and Political Justification for action
~ In its Green Paper, issued in 2002, the European Commission has evaluated the European strategy for energy supply,
~ with attention to priority aspects as growth of demand, supply dependence and meeting international commitments on Climate Change
~ it concluded that nuclear power should remain among these options
~ these principal assumptions are used to extrapolate the current trend up to 2030
~ the conclusions drawn from this scenario are an energy dependence of around 70% and the impossibility of meeting the Kyoto objectives
~ in addition, there is widespread consensus that overall EU policy on sustainable development should include an ambitious strategy on Hydrogen
~ a future sustainable energy system based on hydrogen production and e.g. fuel cells
~ would provide an effective response to the pressing challenges of increased energy demand (especially in the transport sector),
~ security of energy supply, greenhouse gas emissions and air pollution
~ as for the hydrogen production technology without CO2 emission, two practical approaches are available
~ the first is to use electricity to separate water into hydrogen and oxygen by a range of electrolysis technologies
~ the second approach is to use heat directly to drive a thermochemical water-splitting cycle
~ nuclear power could be the basic source of energy for both methods
~ both methods still have a high development potential and feature specific advantages
~ there are good prospects for an early deployment (by the end of the present decade) of a renewed generation of thermal
~ neutron high temperature gas cooled reactors (HTRs) with modular design
~ moreover, in the longer term, the gas cooled reactor technology has a high potential for further developments
~ performance and cost optimisations, use of thorium cycle,
~ deep actinide burning, hardening of the neutron spectrum, etc.,
~ which can improve the competitiveness, open new market areas (hydrogen production, potable water through sea water desalination, industrial process heat and district heating)
~ and lead, step by step, to an improved satisfaction of the goals of sustainable development

Development of the Technology Platform
~ since FP5 preparatory work done within MICANET (Michelangelo Network)
~ opinion of the MICANET Policy Board composed of representatives of the most significant R&D and
~ industrial partners (including new Member States) advising the Commission to focus its R&D efforts in innovative gas cooled reactors
~ many activities at European and international levels, including European co-operation within the Generation IV International Forum (France, UK and Euratom are Members)

Activities (existing and planned in short term)
~ The European High Temperature Reactor Technology Network (HTR-TN)
~ FP5 projects on HTRs and GFRs
~ Proposal for an FP6 Integrated project on V/HTR and a STREP on GFR
~ European participation to the development of VHTR and GFR systems in the context of the Generation IV International Forum
~ Active efforts to bring European and national projects and initiatives together
~ MICANET final report on European efforts to ensure the capacity to develop future reactors

»Japan & Generation IV International Cooperation

Material Science /VHTR

»Nevada RD

»High-Temperature Nickel-Based Alloys for VHTR Applications
~ VHTR is a helium-cooled reactor operating with outlet temperatures exceeding 950°C
~ high yields for energy generation and the ability to supply high-temperature process heat for hydrogen production are the advantages
~ high temperatures impose challenging design requirements on structural materials, particularly for the intermediate heat exchangers
~ nickel-based alloys are the most suitable materials for high temperatures, particularly Inconel Alloy 617 and Haynes 230
~ however, these alloys are not fully ASME code-qualified for nuclear applications
~ basic data are needed to achieve a complete understanding of their behavior at high temperatures
~ researchers must investigate the mechanical properties of these alloys and study surface/subsurface corrosion effects caused by helium impurities
~ as component integrity must be demonstrated over the entire operational lifetime on the order of 100,000 hours,
~ the evolution of properties over time must be taken into account, specifically the effects of thermal aging

Alternatives
~ alternatives in the introduction phase are distribution of liquid hydrogen by ship and truck, and on-site generation of hydrogen

»Nuclear Power 2010 by NRC
~ U.S. now focused on Very-High-Temperature Reactor (VHTR) and Sodium Fact Reactor (SFR)
~ Bilateral projects with Brazil, Canada, Euratom, France, Japan, Korea (I-NERI)
~ Nuclear Hydrogen Initiative
~ Nuclear Energy Research Initiative (NERI)
~ As of 2006, 36 universities are involved in over 70 NERI projects (over $24M)
~ 24 new university research awards in FY 2006 to 17 universities ($10M)
~ 13 - Advance Fuel Cycle Initiative
~ 6 - Generation for IV
~ 5 - Nuclear Hydrogen Initiative