Evolution of nuclear reactors
In the nuclear energy sector, generation III/III+ units represent the current level of BAT (Best Available Techniques). These are the latest nuclear power plant designs, which have better process, safety and economic properties compared to previous generations.
The figure below shows the progressive evolution in the nuclear power sector.
The construction of generation I nuclear power reactors, which put nuclear energy to a peaceful use for the first time, commenced in the 1950s. Following the first power plants, built more for demonstration purposes, the first competitive commercial units – illustrative fig. – nuclear power plant – were put into operation; their construction continued into the 1960s and 1970s. Those power plants are largely out of operation today. The last reactor of this generation that is still in operation is a Magnox (CO2-cooled, graphite-moderated) reactor at the Wylfa Nuclear Power Station, UK.
Construction of generation II power plants commenced in the 1970s. Such power plants currently form the backbone of the nuclear energy sector and their current condition allows extending their operational life beyond the originally planned limits. More than a half of the units are light-water PWR reactors (including the VVER units built in Czechoslovakia and still in operation in the Czech Republic and Slovakia). In addition to the PWR types, the energy industry also makes a wide use of boiling-water reactors (BWR). The Canadian concept CANDU of heavy-water reactors (PHWR) has been employed less widely. British AGR-type and Russian RBMK-type reactors are generation II graphite reactors.
Generation III power plants currently make use of the best available techniques derived from the well-tested generation II types with a number of evolutionary elements. The main differences from generation II are:
- standardised design, reducing the time required for licensing each power plant, required investment costs and construction time,
- simplified as well as more robust design, permitting easier operator work and increasing operating reserves,
- greater availability (90% and more), higher net efficiency (up to 37%), and longer useful life (at least 60 years),
- lower risk of accident featuring serious damage to the active zone (much below 10-5/year),
- higher resistance to external influences,
- potential for better spending of fuel (fuel utilisation increase up to 70 GWd/tU) and reduced amount of waste generated,
- extended period of fuel inside the active zone by using spending absorbers (up to 24 months).
Generation III+ is a direct development of generation III. These reactors have improved operating economics. PWR reactors belonging to generation III+ include EPR units constructed in Olkiluoto, Finland and Flamanville, France or the new Russian AES-2006 reactor (traded as MIR-1200) of the VVER series, Japanese EU-APWRs or AP1000 reactors from Westinghouse.
According to the current state of development, the first power plants of the next generation (generation IV) are expected to commence operation after 2030. They include “fast” reactors that should allow fission of uranium-238 or abundantly available thorium. Several generation IV projects are high-temperature reactors that allow using nuclear energy as a source of heat for industrial facilities and for the generation of hydrogen as an alternative fuel for automobiles.
Safety characteristics of generation III+ reactors
Generation III/III+ projects include new design systems specifically intended to handle selected beyond design basis accidents, such as low-pressure core melting, accidents without shutting down the reactor or blackouts.
The introduction of the new systems for handling beyond design basis accidents or improvement to existing systems (such as increased pressure resistance of the protective envelope, use of double containment for better protection from containment bypass and external impacts) has reduced the probability of active zone meltdown and major leaks by at least one order of magnitude compared to generation II reactors. The hypothetical consequences of design basis accidents on the environment have also been reduced.
Principal safety objective
The nuclear power plant will be designed to fulfil principal safety objectives according to the latest requirements of the International Atomic Energy Agency (IAEA). The principal safety objective is to protect individuals, society and the environment from adverse effects of ionising radiation. To achieve the safety level that is as high as is reasonably achievable, it is necessary to:
- Prevent uncontrolled exposure of persons and release of radioactive materials into the environment;
- Minimize the probability of those events that could lead to loss of control over the reactor core, over fission chain reaction, the radioactive source or any other source of radiation;
- If such events occur, handle the situation to minimize their consequences.
The fulfilment of the principal safety objective will be taken into account in all phases of existence of the nuclear installation, i.e. its planning, siting, designing, manufacturing, fabrication, construction, commissioning, operation up to decommissioning, including transport of radioactive materials and radioactive waste management.
Basic safety requirements
The nuclear power plant will be constructed in compliance with the legislation of the Czech Republic and with the current internationally accepted safety requirements relevant for nuclear technology. Below listed are requirements considered as binding:
- Acts and implementing regulations of the Czech Republic, including international treaties and conventions by which the Czech Republic is bound
- IAEA safety standards (at the level of basic safety principles and safety requirements IAEA SF-1, IAEA SSR) and WENRA safety requirements
The new nuclear power plant will comply, among other standards, with the following radiological criteria based on the latest – illustrative fig. – nuclear plant – requirements of WENRA, WENRA RHWG – Reactor Harmonization Working Group, European Atomic Forum ENIISS – Initiative, IAEA and ICRP:
- authorised limits for radionuclide emissions into the environment will not be exceeded during normal and abnormal operation of the new nuclear power plant; the dose optimization constraint, applying to exposure from releases from all units operated at a site, will not be exceeded for the representative person,
- no event without core melting will result in a release of radionuclides requiring implementation of the protective measures of sheltering, iodine prophylaxis and evacuation of the population anywhere near the new nuclear power plant,
- design measures will be adopted for postulated core-melt accidents to prevent the necessity of population evacuation and the necessity to introduce long-term food consumption constraints; core-melt accidents that might result in early or extensive releases will be virtually excluded.