Advanced Reactor Design and Small Modular Reactors
Abstract
The main class of commercial reactors built from the 1960s to the 1980s is today generally called second-generation reactors. These have been succeeded by a range of third-generation reactor designs, a few of which have been built or are under construction. These reactors attempt to provide lower construction costs, greater efficiency, and use passive safety features to provide the highest level of security possible. A further generation of designs, the fourth generation, are being developed. These encompass a range of different reactor types, including both fast breeder and more conventional reactors, some using high-temperature gas cooling. Alongside these, many companies are also designing small reactors that can be produced in factories before shipping to a power plant site. The aim of these is to reduce the cost of nuclear power and make it more widely available.
Keywords
Third-generation reactors; fourth-generation reactors; nuclear design certification; gas-cooled fast reactor; molten salt reactor; small modular reactor; pebble bed reactor; supercritical water-cooled reactor
The first commercial nuclear reactors that were built in the 1950s and early 1960s, including early pressurized water reactors (PWRs), boiling water reactors (BWRs), and gas-cooled reactors such as the Magnox reactors are considered today to be first-generation reactors. These were followed by a set of classic, standardized designs that were rolled out commercially in many countries. These are now referred to as second-generation reactors.
These second-generation designs include gas-cooled reactors such as the UK AGR and the Russian RBMK but construction of these gas-cooled lines halted in the early 1980s. By the end of the 1980s, construction of all second-generation reactors had virtually halted as cost spiraled and public opinion turned against nuclear power following the accidents at Three Mile Island in the United States and Chernobyl in Ukraine.
In spite of this virtual moratorium on new nuclear plant construction, all the main nuclear plant manufacturers continued to hope for a nuclear renaissance and a spate of new orders. In order to persuade a skeptical public and utilities that nuclear power remained viable, they all developed new reactor designs that incorporated a host of additional safety features as well as new construction techniques to reduce costs. Many of these designs were originated in the 1980s and offered for construction in the 1990s. All were based on water-cooled reactors but only one of these new designs was built before the end of the 20th century. These reactor designs are now called third-generation reactors.
As well as costs rising dramatically during the 1980s as new safety features were demanded, the approval process for new nuclear plants became extended as a result of the introduction of much stricter environmental and safety requirements. Often each proposed construction project had to go through the same, long procedure even if the new reactor was essentially identical to one already authorized for construction elsewhere. In order to make construction simpler and avoid this massive delay, companies developed standardized designs which they sought to have certified in different regions of the world. Approval by a nuclear certification authority meant that, in principle at least, the planning application for construction of such a plant would be streamlined since the actual reactor design had already been approved. This has now become a key strategy for third-generation reactors.
The certification of third-generation nuclear power plant designs is carried out in the United States by the US Nuclear Regulatory Commission. Since 1997 this organization has certified a number of reactor designs for construction in the United States. In Europe, reactors can be certified for compliance with the European Utility Requirements. A number of third-generation plants have been judged to meet these criteria. In the United Kingdom the Office for Nuclear Regulation undertakes generic design assessment.
Meanwhile nuclear design is progressing further still with a range of fourth-generation nuclear reactor concepts. This effort has been led by the Generation IV International Forum (GIF), a collaborative international effort led by the United States which has identified six reactor concepts for potential future development. Alongside this there are other nuclear technologies being explored including high-temperature gas-cooled reactors and a range of modular small reactors that are intended to be cheap and easy to deploy.
Third-Generation Reactors
Third-generation reactors comprise a group of water-cooled reactors that are based on the main branches of second-generation water-cooled reactor design, the PWR, the pressurized heavy water reactor (PHWR), and the BWR. One of the key design improvements in third-generation reactors is the use of passive safety features such that if a reactor goes out of control, naturally driven processes will shut it down. Passive safety means, for example, using natural circulation for the cooling system so that there are no pumps to fail. Other changes are aimed at improving the fuel technology and increasing the overall thermal efficiency of nuclear plants which is low compared to fossil fuel plants. In addition, all third-generation reactors are standardized designs intended to reduce construction costs.
Another feature of third-generation designs, at least for the European market, is the ability to change output rapidly to follow changes in load on the grid. Second-generation reactors were generally designed as base-load power plants intended to operate at close to full output all the time. However the increase in the amount of renewable generation being supplied to grids, especially in Europe, means that many base-load power stations cannot operate at full load all the time because their power is not needed. The European Utility Requirement for new reactors requires that they are able to operate at 25% load and ramp up from 25% to 100% in around 30 minutes. This feature is likely to become important for new reactors on most grids over the coming decade.
Third-generation designs are intended to have an operational life of 60 years, extendable to 120 years with reactor pressure vessel replacement. With improved thermodynamic efficiency, they use their fuel more effectively and the designs aim to burn 17% less uranium for each unit of electricity generated.
The primary third-generation designs were developed during the 1980s but only one was built before the end of the 20th century. That was an advanced boiling water reactor (ABWR) that entered service in Japan in 1996. Since then most companies have refined their designs still further resulting in what are now called Generation III+ reactors, denoting that they are third-generation designs but with additional safety and efficiency features.
The evolution of third-generation designs has led to a bewildering array of reactors available from companies across the globe. The main designs are shown in Table 7.1. While the reactors, and their relationships, are complex, most of them can be traced back to one of the main second-generation reactor designs.
Table 7.1
| Reactor | Manufacturer | Generating Capacity (MW) | Reactor Type |
| ABWR | GE-Hitachi, Toshiba | 1380 | BWR |
| AP600, AP1000 | Westinghouse-Toshiba | 1250 | PWR |
| EPR | Areva and EDF | 1750 | PWR |
| APR 1400 | Korea HNP | 1450 | PWR |
| Hualong One | CNNC and CGN | 1150 | PWR |
| VVER-1200 | Gidropress | 1200 | PWR |
| ESBWR | GE-Hitachi | 1600 | BWR |
| APWR | Mitsubishi | 1530 | PWR |
| Atmea1 | Areva and Mistubishi | 1150 | PWR |
| EC6 | Candu | 750 | PHWR |
| ACR1000 | Candu | 1200 | PHWR |
| VVER-TOI | Gidropress | 1300 | PWR |
Source: World Nuclear Association, Candu.
One of the most successful new reactors in terms of the number operating is the ABWR, designed by GE and Toshiba and based on the US company’s widely used BWR.1 The design has been certified for use in both the United States and in Europe. As already noted, this was the first third-generation design to be built. Four ABWR reactors are now operating in Japan and two in Taiwan. Others are planned. A Generation III+ design that has evolved from the ABWR is the Economic Simplified BWR (ESBWR). The latter has been developed by GE and Hitachi. It further refines the passive elements of the ABWR. The design received certification in the United States in 2014. None has yet been built. There is also a European variant called the EU-ABWR and a UK variant, the UK-ABWR.
While the evolution of the BWR is relatively simple, the heritage of the PWR is complicated by the fact that several different versions appeared around the world. The most direct third-generation descendant of the Westinghouse PWR appears to be the AP600,2 a 600-MW design that was certified for construction in the United States in 1998. However, no reactors were built to this design. Instead, the company—which is now owned by Japanese firm Toshiba—is promoting a Generation III+ evolution of this called the AP1000 with passive features intended to reduce construction costs. The 1250-MW reactor was approved in the United States in 2005. Units based on this design are under construction in China and in the United States. A variant of the AP1000 has also been developed for the Chinese market. Called the CAP1400, it has a generating capacity of 1400 MW. China is expected to approve construction of a plant to this design and the Chinese National Nuclear Corp is also hoping to export the technology.
Another branch of the PWR family is the Japanese APWR. Design of this reactor began in the 1980s. It has been developed by five Japanese utilities in collaboration with Mitsubishi and had support from the Japanese government. Its roots can be traced back to Westinghouse, which was involved at an early stage in the design. This reactor is expected to form the basis for the next generation of Japanese PWR reactors though none has yet been constructed. Meanwhile a US variant, the US-APWR has also been developed.
France adopted the PWR design as the basis for its reactor fleet in 1969. The French design is an evolution of the Westinghouse PWR design but is now considered uniquely French. A third-generation design that takes this a stage further is the European Pressurized Reactor (EPR) which has been developed by Areva, Electricité de France, and the German company Siemens. The design builds on both the French and German national PWR evolutions. Four of the reactors are under construction: one in France, one in Finland, and two in China. The two European projects have experienced severe delays and are not expected to enter service before 2018. The Chinese units, where construction started later, may come into service earlier. The original French PWR design is also the starting point for a Chinese reactor called Hualong One.
Yet another reactor with French heritage is Atmea1. This Generation III + design has been developed by Areva and Mitsubishi and incorporates features of both the EPR and the APWR.
A further third-generation lineage can be traced back to a US company called Combustion Engineering (CE). CE built boilers and steam systems for fossil fuel plants and began to supply steam systems for nuclear power plants in the 1960s. Out of this the company evolved its own nuclear reactor design, a PWR to rival the Westinghouse version. This design eventually merged with the Westinghouse design but before that the line led to the development of its own third-generation reactor design called System 80+. This became the foundation for the Korean APR1400. The APR1400 is based on the Korean Standard Nuclear Power Plant developed by the Korean Electric Power Co. Two APR1400 units are under construction in South Korea and one in the United Arab Emirates. There is a specific European version of this reactor, the APR1400-EUR, which features a double containment and system called a core catcher that will prevent the escape of nuclear material from the core in the event of a melt-down.
The third generation of the CANDU reactor is the ACR series. Two units have been designed, the ACR700 with a generating capacity of 700 MW and the ACR1000 with a capacity of 1200 MW. One of the innovations of the design is to use heavy water in the core but with a light water circuit to extract heat. The reactors would also use uranium enriched to 1.5–2%. The ACR1000 is a Generation III+ design. While none has so far been built, some features of the design have been applied to the Enhanced Candu 6 reactor (EC6) (sometimes referred to as the next-generation Candu reactor—see Fig. 7.1), which is an evolution of the Candu 6 reactor design that was built from the 1980s onward in Canada and elsewhere. EC6 is a 700-MW design PHWR that like its predecessors uses natural uranium as its fuel.
The other two third-generation reactors in Table 7.1 are both evolved from the Russian VVER design. The VVER-1200 is the latest evolution of the standard Russian VVER design to meet Generation III+ standards. It has a gross generating capacity of 1290 MW. Two units are being built in Russia. A further evolution of this design, called VVER-TOI has also been announced. This has a gross power output of around 1250 MW.
Fourth-Generation Reactor Designs and Concepts
While third-generation reactors are beginning to enter service around the world, the industry is already looking beyond these to a new generation of reactors, the fourth generation or Generation IV, which are expected to be ready for commercial rollout from the 2030s onward. One of the key promoters of these new nuclear technologies is the GIF, a collective led by the United States and involving 13 countries including all the world’s main nuclear nations. GIF has selected six nuclear technologies that it believes are the best for future development. Three of these six are fast neutron reactors and only two are dedicated slow neutron designs while a third could be either. Most aim to create a closed nuclear cycle where exhausted fuel is reprocessed to provide new fuel and reduce the amount of waste generated by nuclear generation. The different types are considered below.
Gas-cooled fast reactor: One of the six GIF technologies is a gas-cooled reactor concept using helium coolant in a fast neutron reactor design. The nominal operating temperature is 850°C and the reference unit size is 1200 MW. The core would itself provide the breeding element so there would be no surrounding blanket. Fuel is expected to be uranium nitride or carbide. The primary helium cooling cycle would be used to transfer heat to a secondary helium cycle where high-pressure helium would be used to drive a gas turbine. Heat from the gas turbine exhaust would then be captured to raise steam for a steam turbine, a configuration similar to a combined cycle plant.
Sodium-cooled fast reactor: The sodium-cooled fast reactor has been discussed extensively in Chapter 6. The GIF technology would build on existing fast neutron reactor experience and a range of designs are possible including a core and blanket type reactor as well as a reactor in which the core contains the fertile material that breeds new fuel. Both pool-type reactor cores and more conventional cores with cooling circuits passing through them are possible. Like existing fast neutron reactors, the core temperature would be around 550°C and would operate at atmospheric pressure. Unit size could vary from a small 50–150 MW unit to utility-sized plants of up to 1500 MW.
Lead-cooled fast reactor: Lead or lead–bismuth offers an alternative coolant to liquid sodium for fast neutron reactors, as discussed in Chapter 6. Much of the research into the use of the coolant has been carried out in Russia for nuclear submarine propulsion units but there have been recent designs proposed in the United States and Japan too. The envisaged lead-cooled fast neutron reactor would use a uranium metal or nitride fuel as well as being able to burn fissile waste from conventional water-cooled reactors. The initial design is based around a pool reactor in which the core sits in a pool of coolant. The core would operate at 550–800°C. The reactor could be built in a range of sizes from a 1400 MW utility plant down to small “battery” reactors in which the core is provided with enough fuel to operate for 15–20 years without refueling.
Molten salt reactor: The molten salt reactor is primarily a fast neutron reactor concept in which the fissile fuel is dissolved in the molten salt coolant. The salt is sodium fluoride that circulates through a graphite core, the latter providing an element of moderation. The fissile material is continually added to the coolant and waste removed. A range of fissile isotopes can be used as a fuel, including those recycled from conventional reactor waste. The coolant temperature in the reactor is 700–800°C and the core operates at atmospheric pressure. The molten salt forms the primary coolant circuit with the hot salt used to heat water and generate steam for power generation. There is also a variation of the molten salt reactor that uses a uranium/graphite fuel and core with the molten salt acting only as a coolant. This would be a slow neutron reactor, similar to a gas-cooled reactor but with the molten salt instead of gas as the coolant.
Supercritical water-cooled reactor: One of the drawbacks of the conventional water-cooled reactor is that the steam conditions are relatively mild compared to fossil fuel plants and this leads to low thermal efficiency. The supercritical water-cooled reactor is designed to operate at a much higher pressure than a conventional reactor and at higher temperature in order to raise steam above the critical point of water, the point at which the difference between the liquid and gaseous phases disappears. This allows a simpler steam system design and higher efficiency. Typical core temperature would be 500°C. The design is based on the BWR concept in which steam is raised directly within the core and used to drive the steam turbine. In principle a thermal efficiency of 44% would be possible.
Very high-temperature gas-cooled reactor: The very high-temperature gas-cooled reactor is a slow neutron reactor that operates at 900–1000°C. The core is operated under high pressure, it is graphite moderated, and it is cooled using helium. The hot gas exiting the core can be used directly to drive a gas turbine, then heat recovered from the exhaust of the gas turbine to raise steam for a steam turbine. Unit size is generally small, with generating capacities or around 200 MW or less. High-temperature reactors of this type have been explored since the early days of nuclear development but none has yet reached commercial maturity. The structure of the fuel is a key to this type of reactor. The fissile material, which can be uranium dioxide or a uranium/oxygen/carbon compound, is formed into small particles which are coated with graphite and silicon carbide. The particles are then compressed into larger building blocks that are used to create the reactor core. In one design these building blocks are prismatic blocks, in another billiard ball-sized pebbles. Uranium enrichment of up to 20% is envisaged. With the fissile material locked inside a hard shell, the spent fuel is extremely stable and contained. However the future treatment of the spent fuel presents a difficulty that has yet to be overcome.
Small Modular Reactors
A number of the designs being developed by GIF, above, include small versions. These form part of a general interest in what have become known as small modular reactors. The small reactor concept has been developed in part to try to meet a perceived need to provide a nuclear option for small grids, particularly in developing countries. The standardized, small size is expected to provide an economical means of providing nuclear power while a modular format means that capacity can be added as demand grows by installing additional modules at the power plant site.
Small modular reactors can be based on any of the nuclear technologies including water-cooled reactors, gas-cooled reactors and some of the novel fourth-generation concepts. Both slow neutron and fast neutron technologies can be adapted to meet the demands of a small size. They are generally defined as being 300 MW or less in generating capacity and may be used for both heat and power production. Designs are expected to be simple with many passive safety features. In addition, most of the components should be capable of being built in a factory and then transported to the site, making the cost of construction much lower and the construction schedule shorter.
Small reactors could offer several unusual features. For example, their small size means that they could be sited underground where they would be shielded from accidents caused by external impacts such as an aircraft crash and isolated in case of an internal accident, so that radiation would not be released into the environment. Another proposal is to build “battery” type reactors that contain sufficient nuclear fuel to operate for 10–20 years without refueling.
Interest in these low-capacity reactors can be traced back to the early days of nuclear power development but few have been built except as experimental units. There are a small number operating today. One of the oldest is the Russian EGP-6, a heat and power reactor based on a graphite moderated, water-cooled design that has a rating of 62 MWth and an electrical output of 11 MW of power. Four of these units, which are essentially scaled-down version of the RBMK reactor, have operated since 1976 at Bilibino in Siberia.
Two examples of a small Chinese PWR reactor called the CNP-300 are currently in service, one in Pakistan and a second in China. The generating capacity is 320 MW. In India, meanwhile, a number of small versions of the Canadian Candu PHWR are operating. The earliest of these entered service in 1984 with a generating capacity of 170 MW. More recent versions have an output of 220 MW.
The most important new small modular reactor development is the construction of two 105 MW high-temperature gas-cooled reactors in China. These are based on a pebble bed reactor design which uses fissile fuel embedded into graphite spheres, as shown in Fig. 7.2. The fuel is enriched to 8.5%. The core is cooled using helium which exits the core at 750°C and is used to raise steam at 566°C.
Designs for another 10 or more small modular reactors have reached an advanced stage of development around the world. Many of these are small PWR reactors but there are also small versions of fast neutron reactors too. Most of the research is taking place in the United States, Russia and China, with one project in South Korea.