FAQ - Gen IV Systems Design, Benefits and Challenges

The first generation of nuclear reactor prototypes were constructed in the 1950s and 60s and culminated in the construction of the first series of civil nuclear power reactors. The construction of the second generation of reactors started at the beginning of the 1970s and marked the widespread appearance of Light Water Reactors (LWR), either Pressurized Water Reactors (PWR) or Boiling Water Reactors (BWR), both using normal water as coolant and moderator. The LWRs constructed in the 80s and early 90s are essentially of the same "Gen-II" stock, and these now constitute the vast majority of reactors currently in operation worldwide.
 

More significant evolutionary developments have been integrated into the latest LWR designs available today, especially with regard to design lifetime (typically 60 years compared with 40 years in the past) and safety issues, in particular the behaviour under severe accident scenarios. These new designs are classified as Gen-III, and the first commercially available reactors of this generation are now under construction. In the year 2000, experts from around the world (in the context of the Generation IV International Forum - GIF) began formulating the requirements for a fourth generation of nuclear systems that could respond to the world's future energy needs, in particular increased demand for electricity and reduced CO2 emissions leading to more widespread use of nuclear energy. Making efficient use of uranium natural resources and minimising waste production become major concerns in such a scenario, in addition to satisfying economic competitiveness and maintaining stringent standards of safety and proliferation resistance. The possible designs that are currently under investigation, and around which the GIF organised an extensive R&D international collaboration, have collectively been labeled Gen-IV.

One of the fundamental goals for Generation IV nuclear energy systems is that they will have a clear life-cycle cost advantage over other energy sources. However, since Generation IV reactors are still at an early stage of development and will not be deployed commercially for at least two to three decades, it is difficult to quantify these cost benefits. In particular, Generation IV systems characteristics differ significantly from those of Generation II and III reactors. For example, some designs will have the ability to co-generate industrial process heat and electricity, which will require new models for their economic assessment. This work is currently ongoing within one of the GIF cross-cutting working groups: the Economics Modelling Working Group (EMWG). The "Cost Estimating Guidelines for Gen IV Nuclear Energy Systems" provides detailed information on this topic.

Since they have been designed to operate in a thermal (less energetic) neutron spectrum, current Gen-II and Gen-III Light-Water Reactors (LWRs) can extract fission energy from only a small fraction of the uranium in the fuel (effectively only the "fissile" U-235 component, which makes up less than 1% of natural uranium). Under such conditions, known and easily accessible uranium reserves are capable of sustaining only a few more decades of operation of the world's fleet of LWRs. Four of the six Gen-IV designs currently under investigation are so-called "fast-breeder" reactors, which have the capability of exploiting the full energetic potential of the uranium, thus extending resource sustainability by factors of 50 to 100.

A fast reactor operates in a more energetic neutron spectrum, and is able, via nuclear transformations within the fuel, to "breed" fissile plutonium (Pu-239) from fertile uranium (U-238), which can then be recycled in fresh fuel. In this way, the energetic potential of U-238, representing more than 99% of the original natural uranium, can also be exploited.
 

Gen IV fast reactor designs will represent a radical rethink of current technology. In particular, the objective is to extract and recycle not only the bred Pu-239 but also the other (so-called minor) actinides that are produced in the fuel by nuclear transformation. This "full actinide" recycling will be a key attribute of Gen IV fast reactor systems and associated fuel cycles, as it will reduce the radiotoxicity and heat generation of the ultimate waste for disposal and increase overall proliferation resistance of the fuel cycle.
 

Is it true that the recycling of all minor actinides, along with the Pu-239, will result in much higher levels of radioactivity in many of the processes in the reprocessing and subsequent fuel fabrication plants, and the workforce will therefore need to be adequately protected with appropriate shielding and remote handling apparatus. This technology is already proven in current day facilities dealing with MOX (mixed-oxide fuel) production. In addition, all low-level operational radioactive waste produced in these facilities will need to be managed according to approved and controlled practices.

Recycling all the minor actinides back into fresh fuel enables them to be "burnt" in the reactor and transformed into so-called fission products. These fission products are separated out from the fuel in the reprocessing plant and constitute the "ultimate" waste from the process. This waste must be managed and ultimately disposed of in line with accepted and approved practice. The radioactivity of the waste will decay much more rapidly, and the radiotoxicity is much lower. But once the minor actinides have been removed, it will still be necessary to dispose of the waste in geological repositories to ensure optimal protection of people and the environment over the timescales during which the waste remains a potential hazard. Nonetheless, the quantities of this ultimate waste will be much less than from current fuel cycles for the same energy production. In addition, with the removal of the minor actinides, heat generation is also greatly reduced, enabling a much more efficient use of space in geological disposal facilities.

Along with the physical and administrative monitoring, control and security measures currently in place, careful selection of the fuel composition and reprocessing techniques may further increase the proliferation resistance of the Gen IV nuclear fuel cycle. Making nuclear material less suitable for use in a nuclear weapon, or less prone to diversion for such use, can be achieved in three different ways, which are not reactor but fuel cycle specific:
 

a) By increasing the radiological intensity of the material itself, so that it cannot be handled without severely exposing the people handling it or without heavy and specialized shielding equipment,
 

b) By assuring that at no point during the fuel cycle will the isotopic composition of the fuel be suitable for the production of an explosive nuclear device, without prior complex reprocessing,
 

c) By minimizing the opportunities for diversion, such as during intermediate storage, transport to and from reprocessing, etc.
 

Most of the Generation IV systems involve fast reactors relying on multiple reprocessing and recycling of fuel, which essentially address all three of the above strategies.

You can learn more into the Proliferation Resistance and Physical Protection Evaluation Methodology in the dedicated FAQ and about the GIF PR&PP  Working Group here.

Sodium is highly compatible with the reactor materials, which essentially rules out corrosion problems for the life of the plant. The first reactor to demonstrate inherent safety features that would eliminate the potential for catastrophic accidents like Fukushima was sodium cooled. Sodium is a highly efficient coolant compared to water, meaning that the system can operate at low pressure and high temperature.
 

The main engineering challenge for sodium as a reactor coolant is that it reacts energetically with water, so barriers between the coolant system and the steam system must be robust. Most demonstration and prototype sodium-cooled reactors have encountered water-sodium reactions as a result of design and manufacturing flaws. A second challenge is that sodium is opaque, which increases the difficulty of maintenance and inspection.

You can learn more about SFR here and about GIF SFR System Steering Committee here.

Molten lead is a very heavy coolant that provides advantages for radiation shielding, heat removal, and relative compatibility with the steam system. Lead has also been combined with bismuth to form a coolant with a lower melting temperature coolant, which simplifies design and improves operability. Both concepts would operate at low pressure.
 

Lead presents some unusual engineering challenges. In particular, high-temperature molten lead tends to corrode most metals. While lead-bismuth can reduce corrosion concerns, irradiated bismuth produces polonium, a highly undesirable radioactive byproduct. Lead is also opaque and has the odd characteristic that heavy objects such as nuclear fuel bundles and control rods will float if not secured.

You can learn more about LFR here and about GIF LFR provisional System Steering Committee here.

Molten salt has some interesting benefits in reactor design, with unequaled flexibility. On the plus side, molten salt is an efficient high-temperature coolant whose transparency enables inspection and maintenance of components. The reactor fuel can be dissolved in the salt to allow continuous removal of impurities, or the salt can be used to cool more conventional solid fuel. Both concepts would operate at low pressure. Several salt mixtures have been proposed as reactor coolants, some of which have relatively high melting temperatures, which would complicate keeping the coolant in a liquid state throughout the system. High-temperature salt is corrosive, limiting the materials that can be selected for reactor design. Irradiation of the salt promotes compositional changes that modify the coolant's properties.

You can learn more about MSRs here and about GIF MSR provisional System Steering Committee here.

Within GIF, helium is used as a coolant in two quite different system concepts. Helium has the advantage of being transparent, completely inert, and remains a gas at all temperatures and pressures of interest. Gas-cooled reactors operate at high pressure, but lower than current water-cooled reactors.
 

For reactor designers, the challenge with helium coolant is that its heat removal and retention properties are the weakest of the candidate coolants. For the Very High Temperature Reactor concept, this shortcoming is remedied by providing a large thermal buffer in the form of a graphite (a high temperature carbon material) structure. For the Gas Fast Reactor, structural graphite is not an option, so schemes for pressurized gas flow under all conditions are necessary to ensure safety.

You can learn more about GIF activities of the GFR System Steering Committee here and the VHTR System Steering Committee here. 

Helium shortages can sometimes appear in the current market because it is thin. If a large market for gas-cooled reactors develops, sufficient helium could be captured from oil well production to satisfy the increased demand. Also, helium would be expected to behave like other commodities: short-term supply restrictions would drive up prices, stimulating more exploration and development.

Ordinary water subjected to very high pressure becomes supercritical water, which has a high boiling temperature, greater density, and enhanced chemical reactivity. Supercritical water has been successfully applied in modern coal plants around the world. Its advantages as a reactor coolant are much higher generating efficiency and a wealth of industrial experience that can be applied.
 

The disadvantages of supercritical water are chemical reactivity and a significant variation in density with change in temperature. These properties manifest as corrosion and safety issues for the reactor designer. Supercritical water reactors would operate at much higher pressure than current reactors.

You can learn more about the SCWR system here and about GIF activities on this system here.

In 50 years of nuclear energy development and deployment, the safety performance of nuclear power plants has been continuously improved. Some of these improvements are due to adaptation to state-of-the-art, as occurs with all technologies. Others are the result of lessons learnt following incidents and accidents that have occurred (Three Mile Island 1979; Chernobyl, 1986; Fukushima, 2011) and of the resulting increasingly strict regulatory regime.
 

Generation III systems are equipped with reinforced or double containment buildings and a mixture of active and passive safety systems, which are not only redundant (up to fourfold) and diversified, but also spatially separated. Generation III systems can cope with the consequences of all accidents, including core melt, in a way that ensures the impacts are confined within the nuclear containment and there are no releases to the environment. Over the years, predicted core damage frequencies have been reduced from 0.001 to below 0.000001 (one in a million) per reactor-year; the probability of failure of the last barrier (containment) is another ten times lower.
 

The aim of Generation IV systems is to maintain the high level of safety achieved by today's reactors, while shifting from the current principle of "mastering accidents" (i.e. accepting that accidents can occur, but taking care that the population is not affected) to the principle of "excluding accidents". Generation IV reactors would be equipped with both active safety systems and "passive" safety systems, which rely on natural laws of physics rather than people or machines. These safety systems would be at least as effective as those of Generation III reactors. Certain Gen-IV concepts rely on physical principles which render the most severe accident (e.g. core melt) physically impossible; this is called inherent or intrinsic safety.

Generation IV International Forum Response to the Fukushima Daiichi Nuclear Power Plant Accident

The GIF Experts Group has drafted a position paper on the use of Thorium in the nuclear fuel cycle.