EPR (nuclear power plant)
EPR is the brand name of a series of nuclear power plants. The EPR is a third generation (III +) pressurized water reactor developed by the French companies Framatome (part of the Areva Group between 2001 and 2017 ) and Électricité de France (EDF) as well as the German company Siemens (nuclear division merged with Framatome since 2001) has been. Formerly known as the European Pressurized Reactor or European Pressurized Water Reactor and marketed outside of Europe as the Evolutionary Power Reactor , the abbreviation EPR is today an independent brand name; the long form is hardly used anymore.
Like the Framatome CP series, the EPR has been able to record export orders after the previous P4 and N4 models were only built in France. The first EPR went online on June 29, 2018 in China in Taishan . In 2019, the second block in Taishan went into commercial operation. Three more plants were under construction in 2019: Olkiluoto in Finland since 2005 , Flamanville in France since 2007 and the first of two blocks in the UK, Hinkley Point, since 2018 .
Development history of the EPR
Start of development
Development of the EPR began in 1989 when Framatome and Siemens signed a cooperation agreement to develop an advanced pressurized water reactor . In 1991, Électricité de France and German suppliers also decided to merge their development work. In early 1992, Germany and France published a European Utility Requirement (EUR) for a European Pressurized Water Reactor (EPR) . In 1993 the Reactor Safety Commission proposed to develop common safety standards for future pressurized water reactors. The first two targets were published in February 1994, and the main part followed at the end of the same year.
One focus of the new safety goals was the control of core meltdown accidents . Second generation nuclear power plants did not yet have sufficient safety devices to control a complete core meltdown. The last time there was an accident in France in 1980 at the Saint-Laurent nuclear power plant was a partial meltdown. The accident at the Fukushima Daiichi nuclear power plant in 2011 was also a core meltdown. In order to better understand the behavior of the molten core material, the so-called corium , a number of research programs have been initiated. They created the physical basis for the development of appropriate collection devices for the corium, so-called core catcher ( core catcher ).
Research project COMAS
As part of the COMAS project (Corium on Material Surfaces), the propagation behavior of prototypical core melts was investigated from 1993 to 1999. In this research project, funded by the European Union and the German Federal Ministry of Education, Science, Research and Technology (BMBF), the LAVA code was developed to simulate the spread of the melt. Findings on the spread of lava from volcanology were used and these were supplemented by more detailed models of heat transfer and the rheology of the corium. The validation was carried out in cooperation between RWTH Aachen University and Siempelkamp . In preparation for the large-scale experiments, small laboratory experiments were carried out by Siemens KWU. The test series began with the KATS-14 experiment: 176 kg of oxide mass (85% Al 2 O 3 , 10% SiO 2 , 5% FeO) and 154 kg of iron mass flowed through two channels on cordierite plates in order to validate the propagation speed and temperature profile. This was followed by the actual experiment with COMAS EU-2b: The test mass of 630 kg, known as Corium R , was poured into various channels made of concrete, ceramic and cast iron and measured in the process. To simulate the spread of the melt, the test setup on a scale of 1: 6 corresponded to the EPR version. The composition of the melt consisted of 31.1% UO 2 , 23.8% ZrO 2 , 18.8% FeO, 15.1% SiO 2 , 5.7% Cr 2 O 3 , 4.6% Al 2 O 3 and 0.9% CaO. In all experiments, the necessary heat of fusion was supplied by a thermite reaction .
Research project VULCANO
Further investigations by the CEA took place in France in 1997/1998 in Cadarache with the VULCANO experiments. VULCANO stood for Versatile UO 2 Lab for Corium Analyzes and Observations and was intended to express the versatility of the test series. Compared to the COMAS experiments, which only took place in canals, the flow of the corium from the canal into the spreading area was examined here. The trapezoidal expansion surface was provided with a checkerboard pattern to enable the measurement of the expansion with a camera. Since the spread of the corium in the core trap can be ensured at a high flow rate, the tests concentrated on low flow rates of less than one liter per second. In the initial trials of the VE series, hafnium was used as a uranium substitute to adjust the furnace. It also turned out that the melt was never stopped by crust formation on the front.
| test | Mass percent | Dimensions | Flow rate | Casting temp. | Result |
|---|---|---|---|---|---|
| VE-01 | 50% HfO 2 , 10% ZrO 2 , 10% SiO 2 , 15% Al 2 O 3 , 15% CaO | 12 kg | 0.1 l / s | 2370 K | low spread |
| VE-02 | 70% HfO 2 , 13% ZrO 2 , 7% SiO 2 , 10% Al 2 O 3 | 21 kg | 0.1 l / s | 2470 K | low spread |
| VE-03 | 35% HfO 2 , 5% ZrO 2 , 30% SiO 2 , 25% FeO, 5% Fe | 22 kg | 0.1 l / s | 2420 K | low spread |
| VE-04 | 70% HfO 2 , 13% ZrO 2 , 11% SiO 2 , 8% FeO | 12 kg | 0.7 l / s | 2620 K | accumulation |
| VE-06 | 53% HfO 2 , 10% ZrO 2 , 14% SiO 2 , 13% FeO, 10% Fe | 42 kg | 0.8 l / s | > 2300 K | 45 cm spread |
| VE-07 | 34% HfO 2 , 26% ZrO 2 , 25% SiO 2 , 15% FeO | 25 kg | 0.5 l / s | 2270 K | 55 cm spread |
The first “sharp” experiment VE-U1 with uranium dioxide took place on December 2, 1997. With a composition of 45% UO 2 , 20% ZrO 2 , 20% SiO 2 , 13% Fe 3 O 4 and 2% Fe 2 O 3, the melt corresponded to the corium of the EPR after leaving the reactor pit and melting through the sacrificial plate. The temperature range of the corium corresponded to about 900 K between the solid and liquid components, the temperature at the furnace outlet was kept between 2450 K and 2650 K. A test amount of 47 kg was poured into the sewer at 2.5 kg / s and flowed 1.2 meters down the ramp, with a layer thickness of 2-3 cm. The speed of propagation could be determined to be 10–20 cm / s. Upon closer examination, a very porous structure of the solidified melt could be determined. Further investigations followed. In the VE-U7 experiment, for example, an axially symmetrical partition was placed in the channel and the spreading area. This made it possible to examine two different carrier substrates at the same time. While one half was lined with high-strength concrete , the other half was made of zirconium dioxide . The spread of the melt on the ceramic was examined beforehand using the LAVA code, and good agreement with the experiment was also shown shortly before the corium solidified.
Research project ECOSTAR
Twelve European research institutions and the companies Framatome ANP and Becker Technologies participated in the ECOSTAR project ( Ex-Vessel Core Melt Stabilization Research ). The program was originally supposed to last three years, but after two years a project partner left, whereupon the Karlsruhe Research Center took over the coordination and the program was extended by one year to the end of 2003. Investigations were carried out on the following topics: release of the melt from the reactor pressure vessel (RPV), its spread and the erosion of the concrete floor as well as the cooling of the melt in the core catcher. The following experiments were carried out in detail:
- Spread of the melt: The DISCO experiments at the Karlsruhe Research Center determined the amount of corium that flows into the containment in the event of a high pressure failure of the reactor pressure vessel and does not remain in the reactor pit. For this purpose, a 1:18 model of the EPR reactor pit was examined. The experiments were carried out with water (DISCO-C) and molten, dense metals as a liquid (DISCO-H). Lateral fractures, holes, horizontal slits and tearing of the bottom dome were systematically examined. The hot experiments were carried out with an aluminum-iron melt, steam and a central hole in the bottom dome.
- The KAJET experiments were carried out on jet erosion. In the event of local failure of the reactor pressure vessel under pressure, the corium can be pressed out of it as a compact jet, which accelerates the erosion of the concrete walls. For this purpose, the erosion of concrete by liquid metal jets at gas pressures of 3–8 bar was investigated at the Karlsruhe Research Center. The Corium was simulated using an aluminum-iron melt with a mass of up to 160 kg, which was heated to around 2000 ° C by a thermite reaction. The highest erosion rate could be determined to be 10 mm / s; the depth of erosion in the accident scenario was around 130 mm.
- The transport of the melt was also investigated at the Karlsruhe Research Center. With ECOKATS-1, 600 kg of oxide melt were poured at 2 l / s onto a 3 meter by 4 meter concrete surface. The experiment was used as a benchmark for the propagation codes LAVA, CORFLOW and THEMA. These codes could thus be used for the design and approval of nuclear power plants. At ECOKATS-2 a large mass of 3200 kg oxide / metal melt was poured at 20 l / s onto a 2 m × 2 m concrete surface. This corresponds to flow conditions as expected in an accident scenario. The melt spread 20 cm thick on the surface in less than 60 seconds and gassed out with strong hydrogen flames.
- The phase diagram of the multi-component UO 2 -ZrO 2 -concrete mixture and various oxide mixtures was examined by the CEA and the Czech Nuclear Research Institute. Framatome ANP investigated the interaction of melt and concrete at different heating rates in small experiments. The radioactive decay was simulated by permanent heating. No crust formation could be observed here either.
- The best way to cool the corium in the core catcher was also examined. In the VULCANO tests, especially VE-U7 and VE-U8, the formation of gas bubbles resulted in a rough surface that would improve cooling. However, the melt in the core catcher is five to ten times thicker. If the melt is also cooled from below, the effectiveness of the cooling can be increased by 50 to 600%, as the Royal Technical University of Stockholm found out in the POMECO experiment. In the ECOKATS-2 experiment described above, the 20 cm thick melt was poured with water from above. Despite the blowing out of gases and the moving surface, the flooding process was unspectacular without violent reactions. The melt formed a surface crust on which mini-volcanoes formed; no particles were ejected. The cooling of the melt proceeded slowly, which indicated only a small amount of water ingress. The experiment showed that the melt can only be cooled to a limited extent by flooding from above. If cooling is also applied from below, most of the melt crumbles. This was shown at the Karlsruhe Research Center in three CometPC experiments with simulation of the decay heat : The 800 kg melt solidified in just 30 to 60 minutes, which quenched the processes in the melt . Further investigations were carried out with the DECOBI tests at the Royal Technical University of Stockholm, and a theoretical model for this was developed at the University of Stuttgart .
Completion of development, start of construction of plants
The basic design of the system was defined back in 1998. In 2001 the companies Siemens and Framatome merged their nuclear activities in the company Framatome ANP . This was renamed AREVA NP in 2006 . Work on the EPR was completed under the umbrella of the new company. The technology is mainly based on the experience gained in the construction and operation of the pressurized water reactors of the Konvoi (Siemens) and N4 (Framatome) types developed by the parent companies . Further investigations into the cycle process , as they were necessary for the development of the generation 3 boiling water reactor KERENA in Karlstein am Main at the INKA test stand, were therefore not carried out. From 2003 only individual aspects of the system were developed, so that the development could be successfully completed in the next few years.
In 2005 the building permit was granted for the first EPR at the Olkiluoto nuclear power plant in Finland. With the order, the EPR was able to record its first export success, which France had recently achieved with the power plant types of the CP series from the 1970s and 1980s. The large German contribution to the development of the EPR is also reflected in the Olkiluoto 3 construction site: of the more than 1,600 companies on site (as of 2011), every second was from Germany. In 2007, construction of an EPR began in France at the Flamanville nuclear power plant . Both reactors are currently (2019) not yet operational and have far exceeded their original completion date.
In 2008, the construction of two EPRs began at the Taishan nuclear power plant in the Chinese province of Guangdong . Unit 1 became the first EPR to go into commercial operation on December 13, 2018. Commercial operation of Unit 2 began on September 7, 2019.
In the UK, construction began on two EPRs at Hinkley Point in 2018 .
Generation III +
Extensive research has been carried out at the EPR in order to better counteract a core meltdown. Similar developments also took place in other countries at the same time, so that provisions are generally made for this in all modern power reactors. The countermeasures can be divided into two types:
- Cooling in the reactor pressure vessel ( in-vessel cooling ): The reactor pressure vessel is placed under water from the outside in order to dissipate the decay heat of the corium via the wall of the reactor pressure vessel. Examples are the KERENA from Areva, the WWER-1000 from OKB Gidropress and the AP1000 from Westinghouse.
- Cooling outside of the reactor pressure vessel ( ex-vessel cooling ): A melting of the bottom dome is planned in order to cool the corium in a special device. The advantage over the first method is the cheaper form of the melt, which makes it easier to cool. The disadvantage is the higher effort. Although this device (Engl. By each manufacturer as the core catcher not core catcher ) is called, they still serve the same purpose. A concrete floor reinforced with basalt fibers is used in the ABWR , on which the corium is supposed to spread and is passively cooled. Examples are the EPR from Areva, AES-91 from Atomstroiexport and the ABWR from Hitachi-GE .
In contrast to Generation IV, systems of Generation III + such as the EPR are designed to control core meltdown accidents , but are still dependent on external emergency measures ( offsite emergency response ). Previous nuclear power plants such as the N4 or Konvoi series, for example, require an external water supply to dissipate the decay heat in the event of a complete loss of power supply and loss of the main heat sink. With mobile pumps, water is fed into the secondary side of the steam generator and evaporated. To avoid this procedure, enough water would have to be stored within the system that it can be transferred to a cold, subcritical state and the corium solidifies in the core catcher. The difference between the EPR and Generation IV is due to the increased water inventories compared to the N4 or Konvoi series, but only semantic in nature, since it takes a few hours to warm the water in the flood basin to saturation temperature; As with the AP1000, it takes a few days for complete evaporation. In the EPR, there are almost 2000 tons of water in the flood basin, in addition to the approximately 1600 tons in the EFWS emergency feed system. Due to the design criterion of the EPR, according to which a core meltdown may only have a very limited impact on the environment of the plant, the difference to Generation IV is also negligible. The Risk and Safety Working Group (RSWG) of the Generation IV International Forum has come to the conclusion that the safety standard of the EPR and AP1000 is "excellent" and should be used as a reference for future reactors.
The efficiency of the plants has also been improved. With the EPR, this is mainly achieved through a higher block capacity as well as increased burn-up and process efficiency. Other systems such as the AP1000 try to achieve a cost advantage through simplified safety technology. The reactor technology has also been improved: For example, the EPR can be completely loaded with MOX fuel assemblies at the customer's request , and thorium can theoretically also be added as a fuel. With the ABWR from GE Hitachi, the incubation cycle can be optimized during operation by regulating the coolant flow: At the beginning of the cycle, only a low mass flow is implemented, which increases the proportion of vapor bubbles and generates a harder neutron spectrum and thus increases the conversion rate . In the later phase of the fuel cycle, a higher mass flow is then rolled through the core, which leads to a softer neutron spectrum and the consumption of the plutonium produced. The conversion rate is higher than in older nuclear power plants, but is still below 1. Such reactors are not referred to as thermal breeders , but as upconverters . The EPR's block output is currently still limited by its turbo generator (see below); In the future, a process efficiency of around 39% can be expected. This efficiency should represent the maximum for a conventional cycle; higher efficiencies are only possible with a supercritical steam process . The concepts of pressurized and boiling water reactors are therefore being merged as part of the Generation IV initiative to create the supercritical light water reactor .
Areva
The EPR is the first series of power reactors to be marketed by Areva NP, today's Framatome . In the meantime, the product range has been expanded to include additional systems in order to better cover various customer requirements:
- The KERENA is a further development of the boiling water reactors 72 of units B and C of the Gundremmingen nuclear power plant . The emergency cooling systems work purely passively via communicating tubes . In Karlstein am Main, a complete model of the system was built using the INKA test stand. With a block output of around 1250 MW e and average security technology, the KERENA covers the middle market segment.
- The ATMEA1 is being developed by the 50/50 joint venture of the same name with Mitsubishi Heavy Industries (MHI) and is intended as an inexpensive solution for financially weak customers. This is a kind of heavily scaled-down version of the EPR: the redundancy and circuits have been reduced from four to three, the double containment has been replaced by a simple one, and the emergency cooling system has been simplified. The core catcher was retained, the block output is around 1100 MW e . As of July 2015 there is neither an existing Atmea reactor nor one under construction.
Energy policy of France
France's nuclear power plants are based on four different designs. The first are power plants of the type CP0, CP1 and CP2, which have a capacity of around 900 MW e and were mainly built between 1970 and 1980. Compared to the CP0 and CP1 series, the redundancy of the CP2 series has been increased; from CP1 onwards, water can also be sprayed into the containment in emergencies. These types of reactors were exported very successfully, for example for the Koeberg and Uljin nuclear power plants or the Chinese CPR-1000 reactor series. The following series P4 and P'4 deliver around 1300 MW e output, the Cattenom nuclear power plant belongs to this type. The N4 design was modified from this in Civaux and Chooz with 1450 MW e .
The EPR is the latest series of French nuclear power plants and, according to the will of the Commissariat à l'énergie atomique et aux énergies alternatives, is to replace the previous systems, some of which will remain on the grid until 2050. The EPR can be operated with up to 100% MOX fuel elements and thus as a "plutonium burner". From 2020 - depending on the uranium price - there will be the possibility of using thorium-232 in the breed and feed process. It is expected that up to 27% of the fuel will consist of Th / Pu or Th / U mixed oxides in the core. The EPR is to 2030+ from nuclear reactors Generation IV be supplemented, as the fast breeder reactors , the fission to the possibility of the fast expanding fission (fr. Réacteurs à neutrons rapides, RNR ). The last EPR should then go out of operation before the end of the century, so that electricity will only be generated by fast breeders.
The EPR can replace the older power reactors of the CP series with its block capacity in the ratio of 2: 1. The P4 and N4 series would then have to be replaced by Generation IV reactors, as these plants did not go online until after 1986. It remains to be seen whether this can be achieved in view of the high unit price for an EPR. Also, after the nuclear disaster in Fukushima (March 2011), the public and parts of politics are significantly more critical of nuclear energy, so that it remains questionable whether the EPR will replace existing plants on a large scale.
technology
Circular process
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The EPR is a pressurized water reactor (PWR) with four primary circuits. As is usual with pressurized water reactors, the system consists of a nuclear and a conventional, non-nuclear power plant part. For nuclear part (Engl. Nuclear Iceland ) the double containment of the reactor pressure vessel and the four primary circuits, the building of the emergency core cooling systems and the building with the cooling pond for fuel counted. The conventional part, the turbine island , contains the steam turbine with the generator and the condenser.
The most important components have been adopted in modified form from the N4 and Konvoi series. The reactor pressure vessel is based on the Konvoi series, the steam generator and coolant pumps on the French power plants of the N4 series
The coolant, demineralized water (deionized water), is conducted in the primary circuits under a pressure of 155 bar via the four main coolant lines ( cold leg ) into the reactor pressure vessel, where it flows down the inner wall. At the bottom, the direction of flow is reversed, so that the core with the fuel elements flows through from below and the water heats up from around 296 ° C to around 328 ° C. From there it flows over the four hot-side main coolant lines (Engl. Hot leg ) into the four steam generators (Engl. Steam generator ), which are designed as shell and tube heat exchanger. In the direction of flow behind the steam generators there is a centrifugal pump (main coolant pump, English reactor coolant pump ), which pumps the coolant back into the reactor pressure vessel. In order to be able to regulate the pressure in the primary circuit, a pressure holder is connected to a circuit between the hot side and the steam generator. The mass flow through the core at a nominal load of 4300 MW th is around 23,135 kg / s, 28,330 m³ / h are circulated per circuit.
The primary circuit only has the task of transferring the heat from the nuclear reactor to a secondary circuit, which is designed as a Clausius-Rankine cycle . As a result, there is a phase transition of the working medium. For the sake of simplicity, water is also chosen here. In the four steam generators, saturated steam is generated at 78 bar pressure and around 293 ° C, which flows via four lines at 2433 kg / s each via redundant safety valves from the double containment into the machine house and into the steam turbine, where the turbine generator generates electrical energy. After flowing through the high-pressure turbine, the steam is passed into a heat exchanger, in order to then be fed into the three low-pressure turbines. The steam is condensed into the feed water tank in the six condensers; the cooling water for this is taken from the main heat sink ( ultimate heat sink ), sea or river with cooling tower, at about 57 m³ / s and heated by about 12 ° C. From the feedwater storage tank , the water is fed into the feedwater preheater in two lines with three feedwater pumps - one more is available as a reserve. The water is heated to around 230 ° C in seven stages before it is fed back into the four steam generators and the cycle begins again.
The EPR is marketed with a unit output of approx. 1600 MW e and a thermal efficiency of 37%. These values vary slightly, depending on the average temperature of the cooling water available at the location (sea, river). The thermal reactor output, on the other hand, is a fixed parameter because it forms the basis of all safety analyzes (including possible accident scenarios) and the design of the safety-relevant components. In addition, the performance is also limited by the thermonuclear process itself. The temperature and pressure in the cooling water are precisely matched to one another. If the cooling water temperature were to be increased at the same pressure (by pulling out the control rods or reducing the boric acid concentration in the water), vapor bubbles would form on the fuel assemblies, the moderating ability of the water would decrease and the thermal output would decrease ( negative vapor bubble coefficient ). In addition, the vapor bubbles impair the heat transport into the cooling water and lead to overheating of the fuel assemblies. In practice, therefore, an increase in output usually takes place on the non-nuclear side of the power plant, e.g. B. by increasing the efficiency of the turbine generator. But technical progress can also result in new leeway. So could z. B. on the basis of more precise calculation methods of the evidence that the very conservatively determined design limits of the safety components are not exceeded even at higher performance (temperature and pressure).
Reactor pressure vessel
The reactor pressure vessel is the heart of the power plant, as mass is converted directly into energy through neutron-induced nuclear fission . During nuclear fission, heavy atomic nuclei disintegrate into lighter fission products that have a higher mass defect than the heavy starting nuclide. As a light water reactor , the EPR uses thermal neutrons ; the water in the primary circuit serves as both moderator and coolant. The moderation (braking) of the neutrons increases the cross-section for the nuclear fission of the fuel. When the temperature in the reactor rises, more steam bubbles form, the moderating effect of the water decreases, the cross-section decreases and, consequently, the number of nuclear fission (negative steam bubble coefficient ). From a safety point of view, however, this property hinders any increase in the unit output, so that this can only be increased in the course of the plant's life with better steam turbines and more efficient heat exchangers. A variable amount of boric acid is added to the water during operation. Since boron is an effective absorber for thermal neutrons, the change in the boric acid concentration can compensate for the burn-up of the fuel elements and thus keep the output of the reactor constant. Another automatic power control results from the physical dependence of the reactivity on the fuel and coolant temperature. An increase in temperature in the reactor also means an increased fuel temperature: This increases the tendency of uranium-238, which is difficult to split by thermal neutrons, to absorb neutrons.
The reactor pressure vessel has an internal diameter of 4.885 m and a wall thickness of 250 mm. The bottom dome of the pressure vessel is only 145 mm thick in order to serve as a kind of predetermined breaking point in the event of a core meltdown . With the reactor lid on, the total height is over 12.7 meters, with a mass of 526 tons. The container is made of ferrite steel that is forged into ring-shaped structures and then welded together. The area that accommodates the eight flanges of the primary circuits is forged from a single piece in order to keep the number of welds on the reactor pressure vessel as low as possible. To protect against corrosion, the inside is lined with stainless steel . The water flowing through the four cold sides of the reactor pressure vessel into this flows down along the inside of the wall in order to cool the core from the outside. A flow distribution plate is located on the floor under the guide grille to direct the water uniformly through the reactor core. This has an active height of 4.2 meters and consists of 241 fuel and 89 control rod bundles. After flowing through the upper guide grille, the water leaves the pressure vessel through the four hot sides into the primary circuits. The core is surrounded by a 90-tonne neutron reflector made of austenitic stainless steel, which is supposed to increase the burn-up and breeding factor.
Reactor lid and instrumentation
The lid of the reactor pressure vessel, like the internal structures, is made of stainless steel and is 230 mm thick. It has 89 openings for the control rods, 16 openings for other instruments, four openings for coolant flow measurements and one opening for temperature measurement on the cover.
Since the design of the core instrumentation was adopted from the Konvoi series, it was possible to dispense with openings in the bottom of the container in contrast to the N4 reactors. Of the 16 openings for other instruments, 12 are used for lance yokes . Each of them has three thermocouples (engl. Thermocouple ) for measuring the core exit temperature, six sensors in the core for continuously measuring the neutron flux as well as three to four ball measuring probes , which measure the neutron flux while only batchwise but more accurate and higher resolution. They contain balls made of a vanadium alloy, which are pneumatically blown to the reader after three minutes of irradiation in the reactor. There, the activity of the spheres is typically determined for three minutes at 36 measuring points of a probe in order to recalibrate the instruments for continuous measurement of the neutron flux in the core and to display the three-dimensional distribution of the reactor power. There are a total of 40 ball measuring probes in the reactor pressure vessel. Outside the container there are further instruments for measuring the neutron flux in order to measure the core power and to monitor the sub- criticality during the core unloading and loading .
Reactor core
The fuel elements in the reactor core release heat as a result of nuclear fission and are grouped in rectangular bundles. There are a total of 241 fuel bundles in the reactor pressure vessel, each of which consists of 265 fuel rods and 24 control rod guide tubes in a 17 × 17 arrangement. The bundles have a length of 4.8 meters, an edge length of 213.5 mm and a mass of 735 kg. The bundles are held in shape by ten spacer grids ; the grids also optimize the flow of cooling water around the fuel assemblies. The eight middle grids are made of a zirconium alloy, for reasons of strength the top and bottom grids are made of a nickel alloy . The bottom grid contains a filter to protect the fuel elements from contamination of the primary circuit with particles (due to wear and tear). The top grid contains leaf springs on each side to hold the fuel assemblies in shape against the flow. While the middle grids are connected directly to the fuel assemblies, the top and bottom grids are attached to the fuel bundle with 24 spacers.
The total of 63,865 fuel rods made of an M5 zirconium alloy contain the ceramic fuel pellets made of sintered uranium or plutonium-uranium mixed oxide. The uranium must be enriched to 1.9–3.3% for the initial charge, and 1.9–4.9% during operation. When loaded with mixed oxides ( MOX ), the ratio of 239 Pu to 238 U corresponds to the same energy equivalent as when loaded with light enriched uranium ( LEU ). The EPR can be loaded with any mixture ratio of the fuels, from 100% LEU to 100% MOX everything is possible. Areva puts the burn-up at around 70 GWd / t, the exact breeding factor ( conversion rate ) has not been published. Compared to a convoy system, which reaches around 0.6, this value could be increased with certainty, on the one hand by the steel reflector around the core, which reflects neutrons back into the active zone and thus reduces leakage; on the other hand, the core could be built more compact than the Konvoi series, which is reflected in a higher power density. The dimensions of the reactor pressure vessel are practically identical to those of the Konvoi series, with the EPR having a higher thermal output. While in the Konvoi series around 65% of the thermal energy is released through the splitting of the breeding plutonium, this value could be increased to an estimated 80% in the EPR due to the higher burnup and breeding factor. The data from the CEA and the CNRS only give imprecisely a conversion rate of 0.6 for pressurized water reactors without optimization and 0.9 for EPR for concepts with thorium. Since this value is below 1, the EPR is not a thermal breeder , but, like the Hitachi ABWR, an upconverter . Due to the high breeding factor, the EPR can also develop thorium as a fuel in the breed and feed process. The fuel element U-238 is replaced (in part) by 232 Th. For this purpose, Areva signed a five-year framework agreement with the Lightbridge Corporation on August 3, 2009 to investigate the use of thorium in the EPR, which can be extended by mutual agreement. Tests with Th / Pu fuel elements took place from the 32nd cycle (from 2002) in the Obrigheim nuclear power plant until its decommissioning. The project was led by Areva and the Institute for Transuranium Elements , and part- funded by the European Community .
In order to reduce the excess reactivity in the core, 2 to 8 percent by weight of Gd 2 O 3 are added as a neutron absorber to new fuel elements . The cycle length can be between 12 and 24 months, after which around a third of the fuel elements are replaced with new ones. A typical loading scheme is shown in the picture on the right. The reactivity must also be reduced for control . The fine control is carried out by adding boric acid (H 3 BO 3 ) to the primary circuits. Boron has a high absorption cross-section and reacts in the process
- 10 B + n → 7 Li + 4 He + γ + 2.31 MeV
The proportion of boric acid in the cooling water is reduced in the course of the operating cycle in order to compensate for the decreasing reactivity of the fuel elements together with the decreasing amount of gadolinium. Since one boron atom is consumed per nuclear reaction, new acid has to be constantly added to the circuits, which would make this method of complete control uneconomical. Most of the reactivity reduction is therefore ensured by the 89 control rod bundles, each of which consists of 24 control rods with a diameter of 9.68 mm. The upper 1.34 meters of the control rods are rarely moved into the active zone of the core and contain boron carbide (B 4 C) as an absorber material. The neutron-absorbing AIC metal mixture is located in the lower 2.9 meters of the control rods. This consists of 80 percent by weight silver ( A g), 15% indium ( I n) and 5% cadmium ( C d). The advantage of this composition is the ability of silver to capture neutrons per atom in several nuclear reactions, for example via the following reaction path:
- 107 Ag + n → 108 Ag → 108 Cd + β - + 1.649 MeV
- 108 Cd + n → 109 Cd → 109 Ag + ε + 0.214 MeV
- 109 Ag + n → 110 Ag → 110 Cd + β - + 2.892 MeV
The following cadmium isotopes are stable up to mass number 114 (reactor technology) and decompose to indium , which itself reacts to tin . Since tin has a small absorption cross-section, it cannot be used effectively as a neutron absorber and the silver atom is "used up". The sheaths of the control rods are made of stainless steel, inside there is helium as a protective gas. The majority of the weight of the control bundle comes from the drive, which at 403 kg per bundle is considerably heavier than the 61.7 kg of the bundle itself. The speed of the motor is 375 or 750 millimeters per second, a reactor shutdown occurs with 3 , 5 seconds considerably faster.
Pressurizer
The pressurizer regulates the pressure in the primary circuit. It consists of forged ferrite steel and is double-lined to protect against corrosion. The wall thickness of the 14.4 m high component is 140 mm, the internal volume 75 m³. During operation, the 150-ton pressurizer is filled with 75 tons of water, resulting in a total mass of 225 tons. Like all components in the primary circuit, the pressurizer is designed for a pressure of 176 bar and a temperature of 362 ° C.
The pressure in the primary circuit must always be kept high enough that the cooling medium (water or deionized water ) does not boil in it despite the high temperatures. The pressurizer is only filled with water in the lower part. The pressure is regulated with the help of the steam bubble in the upper part. In order to be able to increase the pressure, there are 108 heating elements in the floor, where the pressurizer is connected to the primary circuit via a pipe. These can partially evaporate the cooling medium in the pressurizer, which increases the pressure in the vapor bubble at its tip and thus also the pressure of the primary circuits. If, on the other hand, water is sprayed into the upper area of the pressurizer, the steam from the steam bubble condenses and the pressure is reduced. Compared to its previous versions (N4, Konvoi), the pressurizer has a larger volume and can therefore better compensate for operational pressure fluctuations.
To protect the primary circuit from overpressure, the pressurizer is equipped with three safety valves, each of which can dispense up to 300 tons of water per hour into a relief tank. In addition to the safety valves, there are two manually controllable pressure relief valves with a capacity of 900 t / h each. The relief tank has a rupture disk through which the water or steam reaches the containment and ultimately the flood tank at the bottom of the containment and can be fed back into the system for cooling.
Steam generator
The four steam generators transfer the thermal energy from the primary circuits to the secondary circuit, the water of which is evaporated here. Each of the 520 t steam generators has a total height of 23 m and is made of ferrite steel except for the tubes of the heat exchanger. The water of the primary circuit flows through the base piece into the 5980 U-tubes made of Inconel 690 . These have an outside diameter of 19.05 mm and a wall thickness of just 1.09 mm. The water flows upwards inside the U-tubes in the feed water of the secondary circuit, and then back down to the bottom piece, and from there into the cold side of the primary circuit ( cold leg ). The feed water of the secondary circuit is fed into the upper area of the steam generator and then runs down the inner walls. At the bottom, the water flows into the heat exchanger, where it is heated by the Inconel tubes. A partition plate is located between the ascending and descending sides of the U-tubes in order to prevent cross flows and to increase the efficiency of the heat exchanger. Only 10% of the feed water flows into the "cold" side of the heat exchanger, where the water flows back down within the tubes. As a result, there is a greater temperature gradient between the feed water and the tube bundles, which increases efficiency. The resulting saturated steam is drawn off upwards and is passed through the steam separator and steam dryer in the upper part of the steam generator; the residual water content is then 0.25%. The emergency feed water system for the Emergency Feedwater System (EFWS) is also located between the steam separator and the feedwater nozzle .
Turbo set
The steam turbine extracts enthalpy from the saturated steam and converts it into rotational energy , which is converted into electrical energy by a generator . The whole complex is called a turbo set . In addition to the reactor pressure vessel, the steam turbine is the second component of the EPR, which was derived from the Konvoi series and not from the N4 power plants. The EPR's turbine generator is the most powerful in the world, with a turbine output of at least 1720 MW e .
The steam is fed into the machine building through the four secondary circuits. Before this is introduced into the high-pressure turbine, it passes the four relief lines with safety valves, via which the steam can be blown off into the environment in an emergency. The twelve-stage high-pressure turbine manages the mass flow of all four circuits alone and releases around 40% of the power, after which the steam is passed through two heat exchangers ( moisture separator reheaters, MSR ). These dry the steam and heat it up again. This is done with the help of superheated steam, which is branched off on the one hand in front of the high pressure turbine and on the other hand from the seventh turbine stage. The heat exchangers also heat the feed water before it flows into the steam generator, for this purpose superheated steam is diverted from the sixth and seventh stages of the high-pressure turbine. The reheated steam then flows into the three low-pressure turbines, which release the remaining 60% of the power, and from there into the condensers. For control reasons, the turbine can also be bypassed with the aid of a bypass.
The turbo set's shaft is 68 m long and is forged from stainless steel . Each of the four rotors is supported by two bearings, so that there are two bearings between the rotors. The blades of the high-pressure turbine are positively attached to the shaft, the connecting piece to the shaft is forged onto it. The low-pressure turbines each use nine stages, the connector is shrunk onto the shaft . The output stages use twisted blades which, depending on the blade length and speed of rotation, bend in a streamlined manner. They are the only blades in the turbine that are not jacketed. The blading of the last stage has a blade length of 1830 mm, the turbine outlet area is 30 m². Since each turbine stage is designed symmetrically, and the steam is thus expanded in both directions along the shaft, a total outlet area of 180 m² is achieved. The turbine housings are double-walled and the individual components are welded together.
The brushless generator converts the shaft's rotational energy, usually 1500 revolutions per minute, into electrical energy. The component is around 17 meters long, weighs 250 tons and is cooled with hydrogen . Compared to air, hydrogen gas has lower gas friction losses and twice the heat dissipation capacity. A multi-stage fan is mounted on the shaft of the generator, which transports the hydrogen gas axially through the windings of the rotor at 5 bar. The warm gas is then cooled in a water-gas heat exchanger and passed through the generator again. The mean temperature is around 40 ° C. The remaining components are water-cooled, the average temperature here is around 45 ° C. The four-pole generator achieves a power factor of 0.9 and an efficiency of around 99%.
The entire complex weighs over 1100 tons. While the first EPR in Olkiluoto was equipped with a Siemens turbine set, the subsequent power plants will be equipped with the Arabelle turbine complex from Alstom. The shaft is welded here, the number of valves in the heat exchanger is reduced and the accessibility of the components is improved. The main difference, however, is the installation of a medium-pressure turbine through which the steam is directed to the heat exchangers before it hits the three low-pressure turbines.
The net efficiency of the system depends, as described above, on the parameters of the cycle. While Siemens speaks of 37% process efficiency, Alstom states an efficiency of over 38% for its Arabelle turbo generator. The turbo set currently represents the restriction of the unit output of an EPR, so that further improvements can be expected here in the next few years. This is also evident from the design service life of just 30 years for replaceable components.
Capacitors
After the three low-pressure turbines, the steam is fed into the condensers, where it is condensed out with water from the main heat sink. To enable efficient condensation, the condensers are operated with negative pressure. The condensers consist of six units, each turbine is connected to two units. The system can also be operated if only one condenser unit is available for each turbine and the other is currently being inspected. The total heat exchanger surface is 110,000 m², one unit weighs 250 tons. For corrosion protection, it is a titanium alloy used. Purification of this costly component takes place in Taprogge process by sponge rubber balls .
The water from the main heat sink is fed through a suction tunnel with an inlet opening of 60 m² to the pumping station. Before it gets there, it is passed through coarse grids, divided into four streams and sucked through fine sieves and chain baskets. In the pumping station, the four flows are pumped to the condensers by four vertical pumps, each with a delivery rate of around 13 m³ / s. To cool all other systems of the nuclear power plant, 4 m³ / s are required, which leads to a total cooling water requirement of about 57 m³ / s. The cooling water is heated by approx. 12 ° C and returned to the main heat sink via the outfall tunnel .
Security technology
overview
Due to the high energy density of nuclear energy, special safety measures are required to allow the release of energy to take place in a controlled manner. Since accidents can never be completely ruled out, further measures are necessary to limit the effects. The applicable safety standard is specified by the responsible supervisory authorities. The requirements vary from country to country: The Mochovce nuclear power plant does not have a containment , which is not an obstacle to operation in Slovakia . The EPR design criterion is based on the requirements of the French and German supervisory authorities from 1993, according to which a core meltdown may only have a very limited impact on the area around the plant. An evacuation of the population should only be necessary in the immediate vicinity of the power plant, and no long-term restrictions on the consumption of locally grown food should be necessary. Further requirements such as earthquake safety are taken from the European Utility Requirements (EUR) .
As in any nuclear power plant, the safety technology in the EPR is redundant and multilayered ( defense-in-depth ). Here are all components that inherently radioactive materials coming into contact (Engl. In the so-called "nuclear island" nuclear Iceland ) summarized. This consists of a decoupled, thick foundation plate made of reinforced concrete ( baseplate ) to withstand a maximum ground acceleration of 0.25 g. The inner containment with the four primary circuits, the In-Containment Refueling Water Storage Tank (IRWST ) and the core catcher are built on this foundation plate in the middle . The primary circuits are selectively connected to the building via shock absorbers and separated from each other by concrete walls. As with the Konvoi series, the inner containment is lined with an approximately 6 mm thick steel liner to ensure additional gas tightness. In the upper part there are catalytic recombiners, which are supposed to limit the hydrogen content in the atmosphere to a maximum of 10% in order to prevent hydrogen explosions (such as in 2011 in the Fukushima Daiichi nuclear power plant ). The wall thickness of the containment made of prestressed concrete is 1.3 meters. The design pressure of the inner containment is given as 5.5 bar at 170 ° C, the containment leak rate (English. Maximum leak rate ) 0.3 vol% per day at the design pressure and temperature. Above that, after a gap, there is another 1.3 m thick containment made of reinforced concrete . There is negative pressure in the 1.8-meter-wide gap, any gases that may escape are extracted and filtered. The degree of separation of the filter is given as about 99.9%.
There are five spatially separated buildings around the inner containment. The four strings of the emergency cooling system are located in the two units facing the machine hall and the units on the left and right of the reactor. In these units, known as safeguard building , all components of the emergency cooling system are housed in quadruple redundancy. In contrast to the N4 series or Konvoi series, where the emergency cooling systems are 2 × 100% or 4 × 50% redundant, the EPR has four times the redundancy with 4 × 100%. The operation of a single emergency cooling line is sufficient to bring the system into a cold, subcritical state. The building opposite the machine hall houses the cooling pool and fuel store. The control room of the plant is located on the floor above the security building facing the machine hall. With the exception of the two security buildings on the left and right of the inner containment, all nuclear facilities are protected against external violence with a 1.3 m thick containment made of reinforced concrete . In addition to the protected building for the cooling pool and fuel store, there are also auxiliary buildings in which, for example, trucks are loaded and unloaded with fuel elements or other radioactive waste.
The emergency power supply is located outside the containment. Two emergency power generators with 7.7 MW each are housed in two spatially separate buildings at some distance from the reactor building, while the emergency power generators in the N4 / Konvoi series are housed in one building. If all four emergency power generators fail, there are two more, each with around 1 MW, distributed across both buildings. These station blackout diesel generators (SBO-DG) can supply the EFWS and LHSI in security buildings 1 and 4 with electricity. The emergency power generators inside are separated from one another by walls. These buildings are earthquake and detonation wave-proof (up to 10 kPa), but unlike the N4 and convoy series, they are not hardened against aircraft impacts. Protection against aircraft impacts is only provided by the spatial separation. Every emergency generator has enough fuel to run for up to 72 hours without refueling.
The so-called residual risk, i.e. the multiple failure of various safety systems due to technical defects without external influence, is determined with the help of a probabilistic safety analysis, as in aviation. The highest permitted probability of a core damage frequency (CDF ) and the probability of the release of a large amount of radioactive material ( large release frequency, LRF ) are specified by the supervisory authorities. For example, Areva and EdF give a probability of 6.1 × 10 −7 / year for a core meltdown accident for the EPR , which is below the value specified by the European Utility Requirements (EUR) . The maximum value here is 10 -5 / year, and 10 -6 / year for a core meltdown with limited health effects on the environment. Compared to the protection of the plant against flooding, this value is very low: In Great Britain, for example, a nuclear power plant may be flooded once every 10,000 years (1 × 10 −4 / year). A somewhat lower value of 1: 14,000 years is currently being considered (7.1 × 10 −5 / year). The likelihood of an incident is heavily dependent on external factors and human error.
Subsystems
The safety concept of the EPR contains various subsystems at different positions in the power plant, which are listed here. These systems are located in the inner containment:
- In-Containment Refueling Water Storage Tank (IRWST) : The flood tank is located in the EPR within the containments and holds 2000 t of boric acid mixed water. In contrast to the N4 / Konvoi series, there is no need to switch from pressure accumulator feed to core flooding and circulating operation.
- Core-Catcher (CC) : The core catcher of the EPR has a spreading area of 170 m². Strictly speaking, the core catcher consists of an aluminum plug with a concrete cover under the reactor pit, the flow channel into the spreading area and the spreading area itself. The walls of the reactor pit and the flow channel are made of zirconium dioxide and have a concrete cover up to 50 cm thick as a sacrificial layer (engl. sacrificial concrete provided). The floor and the side walls of the spreading area are made of cast iron and are provided with a 10 cm thick sacrificial layer of concrete. When the corium arrives in the core catcher, two pre-tensioned control wires are destroyed by the heat (for reasons of redundancy) and the IRWST water is conducted under the spreading surface via communicating pipes . There the core catcher flows through cooling channels from below, and the water then flows into these via the side walls. The evaporating water is fed through a chimney into the upper area of the containment.
- Filtered pressure relief ( venting ) : In the event of certain malfunctions (see below), the evaporation of water leads to a pressure increase in the containment. Normally, spraying cold water into the containment would release the pressure. If the spray system is not available due to a fault or the failure of all diesel generators, pressure relief to the atmosphere is possible. With the help of various filter systems, most of the radioactive substances are retained (exception: noble gases). However, this system is not a standard at EPR, but is only installed at the customer's request. A venting system is currently only planned for the EPR in Finland.
The following system components are located in the four security buildings:
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Safety Injection System / Residual Heat Removal System (SIS / RHRS) : The system for the safety feed into the primary circuits is also responsible for the removal of decay heat after the system has been shut down. Distributed across all four security buildings, an SIS / RHRS is assigned to each primary circuit. The system consists of two components, one each for high and low pressure feed. There is therefore 4 × 100% redundancy at every feed pressure.
- Medium Head Safety Injection System (MHSI) : The medium pressure
- Low Head Safety Injection System (LHSI) : If the pressure in the primary circuits drops to 20 bar or lower, the emergency cooling system switches to the low-pressure injection system. This system also sucks in water from the flood basin (IRWST) and conveys it via a heat exchanger into the cold line of the reactor pressure vessel. After a loss of coolant accident, water can also be fed into the hot line. The heat exchanger of each LHSI is connected to the Component Cooling Water System (CCWS) , which dissipates the heat to the main heat sink. In the case of longer idle times or a fuel element replacement, the decay heat is removed with the help of the LHSI-CCWS-ESWS circuits, which is why these circuits are also referred to as Residual Heat Removal System (RHRS) .
- Control of the water inventory in the primary circuit by injecting and draining water.
- Control of the boric acid content in the water, and thus the reactor output. For this purpose, either boric acid or demineralized water is fed into the circuit. As the cycle length increases, the boric acid concentration in the water is reduced to almost zero.
- Control of the nitrogen gases dissolved in the water, as well as the blowing off or blowing in of these gases. As a rule, the water in the flood basin (IRWST) is treated with it.
- Chemical substances can be added to the water in the containment for water treatment.
- Treatment of the water in the primary circuit, such as cleaning, outgassing, storage, etc.
- Supply of the seal of the coolant pump of the primary circuit with cooling water, collection of the leakage water .
- Supply of the primary circuit with the necessary boron solution to bring the system into a cold, subcritical state.
- Auxiliary spraying of water into the pressurizer to condense the steam and lower the pressure in the circuits ( auxiliary spray ). This task is normally taken over by systems that belong to the primary circuit; this is only carried out by the CVCS in order to create start conditions for the SIS / RHRS or if the dedicated system fails.
- If there is a small leak in the primary circuit, the leakage water can be replaced by the CVCS.
- Feed-and-bleed cooling through water injection.
The following systems are protected by the outer containment and are located opposite the control room:
- Emergency boron system (EBS) : The emergency boron system is used if the reactor emergency shutdown should fail. For this purpose, boric acid is pressed into the reactor pressure vessel at a pressure of up to 260 bar using two lines with three pumps in order to make the reactor subcritical . Because of the two strings, the system has 2 × 100% redundancy.
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Fuel Pool Cooling and Purification System (FPCPS) : The cooling and cleaning system of the cooling pool consists of two systems:
- Fuel Pool Cooling System (FPCS) : Discharges the decay heat from the cooling pool to the CCWS and is designed with double redundancy; two pumps are available for each of the two cooling lines. For emergencies there is also a third cooling line that has only one pump and can also transfer the heat to the CCWS.
- Fuel Pool Purification System (FPPS) : Consists of several circuits that purify the pool water above the reactor, the water in the cooling pool and the water in the flood pool. The system can also pump water into or out of the pool above the reactor or the cooling pool.
The following systems are located outside the containment:
- Essential Service Water System (ESWS) : The system is housed on the site of the power plant and has four-fold redundancy, with one line being assigned to each security building. The system takes over the heat from the CCWS through heat exchangers and transfers it to the main heat sink. Two lines also have heat exchangers within the flood basin (IRWST) so that it can also be cooled.
Grace period
As mentioned above, in certain accident scenarios, nuclear power plants require an external water supply in order to dissipate the decay heat. The time between the occurrence of the accident and the need for an external water supply is referred to as the waiting period. How much time is available to the staff and the company fire brigade for this depends on the type of system: In the case of the boiling water reactors of construction line 72, for example, the water inventory of the feed water tank is passively fed into the reactor pressure vessel so that the reactor is cooled for at least 2 hours . During this time, an external water supply must be built up, otherwise the core will lie dry and melt after this time has elapsed . In modern nuclear power plants, this period has been increased significantly in order to ensure that feed-in is carried out on time even under adverse conditions.
Except for the core catcher for a core meltdown accident, the EPR mainly uses active safety systems. B. With the AP1000, more attention was paid to passive systems. Passive systems do not require any external energy supply. The triggering of a passive safety component is often irreversible and must not take place at the wrong time or in the wrong situation. Active systems, on the other hand, enable targeted adaptation to the respective situation, but require electrical energy to function. A complete failure of the electrical energy supply has a stronger influence on the waiting periods.
The waiting times of the various accident scenarios can be estimated by the approximation formula for the decay heat over time integrated is and set equal to the heat capacity of the water inventory. As a conservative assumption, only the heat of evaporation of the water of 2.26 MJ / kg is taken into account, pressure and heat capacity are not taken into account. A core output of 4500 MW th and an 11-month fuel element cycle is expected. The water inventory of the emergency feed water system is assumed to be 1600 t (real 1634 t), a possible water supply from the fire extinguishing system is ignored. In core meltdown scenarios, only the delay caused by the sacrificial plate is taken into account; the time required for the core to melt together and the bottom dome to melt through is also ignored.
- Normal shutdown of the plant : In this case the reactor becomes subcritical due to the retraction of the control rods. The decay heat is now released via the steam generator to the secondary circuit, where the turbine is shut down, and then released by the condensers to the main heat sink. From a steam generator temperature of less than 120 ° C, this path is no longer available and the pressure in the primary circuits is reduced via the pressurizer until the Safety Injection System / Residual Heat Removal System (SIS / RHRS) can start. At a pressure of 85 to 97 bar, the medium pressure system is used to feed in, below 20 bar the low pressure system. Both draw the cooling water from the flood basin (IRWST) and feed it into the reactor pressure vessel. The water leaves the primary circuits via the lines of the Chemical and Volume Control System (CVCS) back into the flood basin. Only when the low pressure system (LHSI) starts is the decay heat transferred to the main heat sink via the LHSI-CCWS-ESWS circuits. The energy for this is obtained from the public grid.
- Breakage of a main coolant line : This is the design accident for N4 and convoy systems and is referred to in English as a loss of coolant accident (LOCA) . In the EPR, a break in a main coolant line leads to a reactor shutdown and the start of the pumps of the Medium Head Safety Injection System (MHSI) . The feed can be started from a delivery pressure of around 40 bar, as the pressure difference to the feed pressure of around 90 bar can be compensated for by the 4 collecting tanks. With increasing emptying of the collecting tanks, the full feed pressure is reached by the pumps, while the pressure in the circuit continues to drop due to the leak. As in the case above, the LHSI takes over the supply and heat removal from 20 bar. Due to the flood basin in the containment, there is no need to switch from pressure accumulator feed to core flooding and circulation. It does not matter whether the water leaves the primary circuit via the CVCS, the pressurizer or the break point. The energy for this is obtained from the public grid.
- Loss of coolant accident and loss of the external power supply, only one emergency power generator starts : This case is beyond the scope of design for convoy systems. However, it must be taken into account that the N4 series has 2 × 100% redundancy, so the failure of 3 emergency power generators is only possible with convoy systems. In the case of N4 nuclear power plants, only one emergency power generator would be available (controllable) or none (beyond the scope of the design). The EPR does not change anything compared to the above scenario, only the energy is provided by the remaining unit and only one security building works with 100% cooling capacity.
- Failure of the main heat sink : In this scenario, the decay heat can no longer be dissipated after the reactor shutdown . In this case, feed water is pumped into the steam generator, evaporated and discharged outside. From a conservative point of view, only the water that is already in the steam generator is available here. According to Areva, if the feed water pumps fail, it will take at least 30 minutes before the steam generator runs dry. The Emergency Feedwater System (EFWS) is then used: a total of 1600 t of water is fed into all four steam generators in order to evaporate there and to be discharged outside in front of the turbine. During this time, the plant fire brigade can connect to the suction valves of the EFWS water tank and mobile pumps in order to start the emergency feed after the EFWTs have been emptied. If an emergency feed-in is not possible by then, the primary circuits are depressurized and the water from the flood basin is rolled through the core. About 12 hours after the start of evaporation of the IRWST water, venting through a radionuclide filter is necessary, which is why the vicinity of the power plant must be evacuated. If, after 141 hours, it is still not possible to pump water into the flood basin via external sources in order to cool the system using the feed-and-bleed method, an beyond-design-basis incident occurs, as the core meltdown then occurs and the core catchers are no longer cooled can.
- Failure of power supply and main heat sink : This is a design basis accident in Finland . The Finnish supervisory authority STUK demands that the fuel elements do not suffer any significant damage. If personnel measures are necessary to stabilize such a situation, sufficient time must be available and the suitability of the method must be proven. In this case, after the reactor shutdown, the water that is already in the steam generator will only evaporate for 30 minutes. Since there is no electricity available for the EFWS after this time, some of the staff have to leave the control room and start the two emergency diesel units manually in order to feed the water inventory in security buildings 1 and 4 into the assigned steam generator. Since the valves between the EFWTs are opened, the entire water inventory is also available here. Water tanks and mobile pumps can be connected to the suction valves of the EFWS in order to start the emergency feed after the EFWTs have been emptied. Obviously, this time was sufficient for the STUK, so that the EPR in Finland received the design license. For comparison: The preparation and implementation times required in German systems are around 2 hours. If an emergency feed-in is not possible by then, the primary circuits are depressurized and a core meltdown occurs. After the reactor pressure vessel has melted through, the corium remains in the reactor pit for about an hour until the “stopper” melts and the melt pours into the core catcher. The heat destroys the pre-tensioned control wires and the IRWST water is conducted under the spreading surface through communicating pipes to cool the core catcher and corium. About 12 hours after the start of evaporation of the IRWST water, venting through a radionuclide filter is necessary, which is why the vicinity of the power plant must be evacuated. If, after a certain time after the start of the accident, no water can be pumped into the flood basin via external sources, an beyond-design-basis incident occurs because the core catcher can no longer be cooled.
- Breakage of a main coolant line and failure of the power supply : In this scenario, the water inventory of the steam generator and the EFWS is not available because the primary circuit is depressurized due to the leak. Optionally, the main heat sink can also no longer be available, which would not change the effects. In this case, a core meltdown is inevitable and the waiting period is the lowest, as only the 2000 t of water in the flood basin can dissipate the decay heat. Here, too, after the reactor pressure vessel has melted , the corium remains in the reactor pit for about an hour until the sacrificial plate ( melt plug ) melts and the melt pours into the core catcher. After about 13 hours, venting through the radionuclide filter over the 100 meter high exhaust air chimney is necessary, which is why the immediate vicinity of the power plant has to be evacuated. If no water can be pumped into the flood basin after this time has elapsed, an beyond-design-basis incident occurs because the core catcher can no longer be cooled. In all core meltdown scenarios, the situation can be stabilized if at least one of the outer security buildings can resume operations to cool the flood basin. The heat is transferred to the main heat sink via the CHRS-CCWS-ESWS circuits or directly via the Essential Service Water System (ESWS) . The Containment Heat Removal System (CHRS) will spray water in the upper area of the containment to partially condense out the steam, which lowers the pressure in the containment. Radioactive particles are also washed out, which reduces the radiation exposure of the environment during venting. If the corium in the core catcher falls below the evaporation temperature, the water is no longer sprayed into the containment, but pumped directly into the core catcher in order to completely flood the reactor pit and core catcher.
| BE change | Start of cycle | End of cycle | |
|---|---|---|---|
| Normal cooling pool (1486 m³) | |||
| Warming to 97 ° C | 4 h | 13.6 h | 35.3 h |
| Top edge BE | 33 h | 107 h | 272 h |
| Cooling basin with pipe leak (1195 m³) | |||
| Warming to 97 ° C | 3.3 h | 11.1 h | 28.9 h |
| Top edge BE | 32 h | 105.9 h | 266 h |
The spent fuel pool is located between the inner and outer containment and must also be cooled. Since the waiting period in the event of a power failure depends heavily on the loading of the basin, the information provided by Areva and EdF to Nuclear Technologies , which is conducting the peer review in the United Kingdom as part of the Independent Nuclear Safety Assessment (INSA), is quoted here . Here, the reactor output is assumed to be very conservative with 4900 MW th and waste heat from the cooling basin of 22.3 MW during loading, 6.8 MW at the beginning of the cycle and 2.7 MW at the end. The information has a safety margin of 15%. The cycle is assumed to be 18 months, with the loading being 100% with MOX fuel elements. The waiting period when replacing fuel elements is purely theoretical, since maintenance and repair work is also carried out in the power plant during this time. The lower case is based on a pipe leak in the Fuel Pool Cooling System (FPCS) , which reduces the water level in the cooling pool.
The waiting times of the decay basin are comparable to those of the nuclear reactor. The emergency feed in the event of a power failure occurs here via the fire extinguishing system. If electrical energy is available, the water can also be circulated between the flood and cooling pools via the Fuel Pool Purification System (FPPS) . Since the cooling basin is located between the inner and outer containment, radioactive gases escaping into the air gap are sucked off and filtered.
The waiting times of the EPR for various accident scenarios are on average depending on the situation: An AP1000 from Westinghouse has an IRWST with 2236 m³ of water, with a nominal thermal output of 3400 MW. Westinghouse itself specifies a waiting period of 72 hours. The limitation lies in the containment cooling system, as the water on the roof of the system has evaporated after 72 hours (3 days) and has to be topped up by pumps. This means that the power plant should get by for at least 30 days without a main heat sink. Without container cooling, it takes about 24 hours to reach design pressure and venting is essential. The fuel elements in the spent fuel remain covered by water for up to 72 hours without cooling. GE Hitachi's ESBWR can also be passively cooled for up to 40 hours before the water has evaporated in the Passive Containment Cooling System (PCCS) .
Control technology
Around 150 to 300 people work in the power plant during power operation. The system is controlled from the control room , which is located above the two security buildings that face the turbine building. The control center is thus protected from the nuclear cycle by the double containment and from external violence by the external containment. The control technology is digital and derived from the N4 series. It is divided into three so-called "levels": Level 0 includes all switches and sensors, Level 1 the reactor control and safety systems, Level 2 the user interface . The user interface is connected to the subsystems via a bus system , with all connections being redundant and diverse . For example, the emergency cooling systems and the EFWS each have four independent control systems. Teleperm XS from Areva NP is used as the control technology. To display the system status and incident management, Teleperm XS has a special Qualified Display System (QDS), which allows the reactor operator to intervene and control in the control room within certain limits. The system uses Intel Pentium M processors on motherboards in AT format . About PS / 2 interfaces , for example, a keyboard entries are made. The widgets of the graphical user interface are created on a computer with openSUSE that contains the QDS development environment. The download takes place via Ethernet when the computer starts up. Another Linux PC with the QDS Service Unit must be available for installation , where a user can select the desired program. The computer is also responsible for the handshake , the monitoring of the download and the self-tests as well as for the recording of all screen activities. It is not required for ongoing operations.
In a joint statement, the supervisory authorities of Finland, France and Great Britain criticized the design of the control technology, as they believe that the safety systems for control in the event of exceptional events were very closely interconnected with the control system for normal operation. The wishes were then fulfilled by the manufacturer consortium, so that the Health and Safety Executive (HSE), as the parent authority of the Nuclear Installations Inspectorate (NII), approved the control technology at the end of 2010. The sensors and switches in the emergency room are now supplemented by an analog hardwired backup system (HBS). According to previous plans, the US EPR will continue to be controlled purely digitally, the Nuclear Regulatory Commission has not reported any concerns in this regard.
The EPR's load following capability, important for power plant management , is specified as follows:
- In the peak load range between 60% and 100% of the nominal load with 5% / min during 80% of the fuel cycle
- In the low load range with 25% to 60% of the nominal load with 2.5% / min during 80% of the fuel cycle
- You can drive for up to two days at medium power without losing flexibility; only then does the load-following ability decrease
- At medium power, the momentum reserve of the turbo set can contribute to the load following ability:
- One step of 10% of the nominal power with a ramp of 5% / min
- A ramp with 10% / min for a short power cycle
Since only as much electrical energy can be fed into an electrical distribution network as is needed by the consumers, the momentum reserve of the turbo set is used to control small fluctuations. Since nuclear power plants in France also cover the medium load range, France has one of the largest pipeline networks in Europe, so several power plants can jointly compensate for fluctuations in demand.
Versions
Standard EPR
The standard EPR is the version that was originally developed for Germany and France. France wants to use it to replace its older CP series nuclear power plants. The Italian utility Enel signed an agreement with EdF on November 30, 2007 to participate in the construction of six EPRs in France, with Enel taking a 12.5% stake. In addition, the Italian government under Prime Minister Berlusconi planned to build four to five own EPRs. However, this was prevented by a referendum. The first EPR in France is being built at the Flamanville site. The technical details and problems when building the version are described above.
France :
-
Flamanville Nuclear Power Plant
- Unit 3 (start of construction 2007, planned commissioning 2020)
-
Penly Nuclear Power Plant
- Block 3 (abandoned construction project)
-
Taishan Nuclear Power Plant
- Block 1 (construction started in 2009, commercial operation since December 13, 2018)
- Block 2 (construction started in 2010, commercial operation since September 7, 2019)
India :
-
Jaitapur nuclear power plant
- Block 1 (planned start of construction 2013, construction project 2015 abandoned)
- Unit 2 (planned start of construction 2013, construction project 2015 abandoned)
In March 2015 Areva announced that it would pursue a new business strategy. This does not involve building new reactors. This also applies to the reactors in Jaitapur.
FIN-EPR
Based on studies by the Technical University of Lappeenranta (LUT), according to which electricity from nuclear power is the cheapest solution, the energy supplier Teollisuuden Voima Oy (TVO) applied for a new building in November 2000, which was approved by the Finnish Parliament in May 2002. TVO then selected the Areva EPR. Since March 2007, two more nuclear power plants have been tendered, which were approved by the Finnish Parliament in July 2010. The main difference between the FIN-EPR and the standard version is the reduced burn-up of 45 GWd / t.
Finland :
-
Olkiluoto Nuclear Power Plant
- Unit 3 (start of construction 2005, planned commissioning March 2021)
UK EPR
The UK gets around 18% of its electricity needs from nuclear power plants and has been planning a significant expansion since 2006. The companies Électricité de France (EdF), Horizon Nuclear Power and NuGeneration were able to bid for building lots at eight different locations. A total of around 19 GW of generation capacity is planned, which corresponds to a doubling of the share of nuclear power. On July 18, 2011, the UK Parliament approved the largest new nuclear power plant construction program in Europe. EdF applied for planning permission for the Hinkley Point C nuclear power plant just ten days later, after the municipality approved the construction project. The building permit was issued in March 2013.
However, due to economic considerations, the construction is unsafe. Construction costs of around 16 billion pounds (approx. 19 billion euros) are planned for the two reactors with a total output of 3650 MW, which means that construction without state subsidies for EdF would not appear economically feasible. Therefore EdF negotiated with the government about a guaranteed electricity purchase price . An agreement was announced in October 2013, but the EU's approval is required due to the subsidization of the project. In order to make the project profitable, the British government guaranteed a feed-in tariff of 92.5 pounds / MWh plus an annual inflation adjustment (currently 103 euros / MWh) for 35 years. This is almost double the average UK electricity price. This is roughly equivalent to a subsidy of £ 4 million per day, or £ 50 billion over 35 years.
The nuclear reactors should go online in 2023 and are expected to be operated for 60 years.
-
Hinkley Point C Nuclear Power Plant
- Block 1 (start of construction March 31, 2017, planned construction period 10 years)
- Block 2 (start of construction unknown)
-
Sizewell C nuclear power plant
- Block 1
- Block 2
US EPR
Because electricity companies in the United States were planning to build new nuclear power plants, Areva applied for design certification of the EPR by the Nuclear Regulatory Commission on December 11, 2007 . Possible building sites are Nine Mile Point , Bell Bend , Calvert Cliffs and Callaway . However , at Nine Mile Point, Callaway and Calvert Cliffs, the application for a combined license (COL ) was suspended at the request of the operators. The project for the new Bell Bend nuclear power plant was canceled in September 2016 after Areva suspended US EPR certification. The main difference between the US-EPR and the standard version is, besides the generator, the higher stability up to a ground acceleration of 0.3 g.
construction
Approval process
The construction of a nuclear power plant is time-consuming and costly. In addition to an investor, usually the energy supply company, a design license for the reactor model issued by the national authority for reactor safety as well as political approval for the construction of the plant are required for the construction. The design license can either generally be granted to all systems of one type, as is the case in Great Britain, or depending on the construction site. The American Nuclear Regulatory Commission (NRC), for example, starts with the Acceptance Review , in which basic things such as schedules etc. are agreed. This is followed by the safety review , in which it is checked whether the design meets the safety requirements of the legislator. The subsequent environmental review is location-dependent and takes into account water temperatures and defines other limits necessary for operation. This is followed by a public mandatory hearing of the residents, which is usually followed by the issuing of the building and operating license. The whole process takes about six years. In Great Britain, the process is divided: Here, the reactor design is first examined in general to determine whether it meets the safety requirements of the legislature, and then the design license is awarded. In order to obtain the building permit, however, a separate application must be made, in which the building site is checked for its suitability, which takes around 18 months. The political approval also differs from state to state: While in France and the USA the general permit to build a nuclear power plant is sufficient, in Finland every new building must be approved by the Finnish parliament . In the UK, both the UK Parliament and the local community must approve the construction.
Construction process
Once the administrative hurdles have been overcome, which can take over six years, construction of the EPR can begin. The earthworks to prepare the construction site take about a year. During this time, the tunnels for the cooling water supply and discharge (sea or river water) are dug. Officially under construction, according to the guidelines of the International Atomic Energy Agency (IAEA), the power plant is only (Engl. With the introduction of the first concrete first concrete ) in the casting of the foundations and the establishment of the nuclear power plant unit.
After the completion of the base plate of the nuclear power plant section (English nuclear island ), the construction of the inner steel liner begins. Ring segments are welded on the construction site, which are stacked on top of one another on the base plate with cranes and welded together. A ring has a diameter of 42 meters, a height of around five meters and a total weight of 218 tons. Since the steel liner was taken over from the Konvoi series and is not available in French power plants, there were several delays in its production. At the same time, the inner and outer containment are being built, with the steel liner always being pulled up in advance. Before the containment dome is put on, the internal bridge crane ( polar crane ) must be used. The dome is also welded together on site at the construction site, set into position with a crane and welded there by hand. The dome has a circumference of 147 meters, a weight of 270 tons and is the largest to date for a nuclear power plant. Then the inner containment made of prestressed concrete is completed first and the concrete construction of the containment with the outer containment made of reinforced concrete is completed. At this point in time, more than 100 km of cables had already been laid. The installation of the components begins while the containment is being completed. The pipelines for the security systems such as the Essential Water Service System are being laid and the installation of the emergency batteries is being started. For this purpose, there is a circular opening between the containment and the environment through which the components can be reached. The reactor pressure vessel as well as the steam generator and the pressurizer are also passed through these. The components are lifted by a construction site crane and placed on a rail vehicle. This runs on tracks that extend from a scaffolding attached to the building through the circular opening to the inside of the containment. Inside, the components are lifted up again by the overhead crane, placed in their position and installed there. After the steel dome has been installed, the installation of the emergency power generators in the corresponding buildings will also begin. The nacelle with the turbo set will be built parallel to the nuclear power plant section and is almost completed at this point. Next, the flood basin (IRWST) is filled with water on a test basis and this is drained off again. Now the labor-intensive degreasing of the liner begins: dust and grease are washed away with an aqueous solution and then the surface is passivated . After rinsing with demineralized water, the liner is examined for remaining corrosion spots, which are then mechanically removed. In the meantime, the overhead crane of the fuel storage facility is put into operation so that all the installations can be carried out here as well. If the domes of the two containments are also ready, the chimney is placed in one piece on the building and fastened there. After the construction work has been completed, the interior fittings are installed, the cables are laid and other fittings are installed. For example, the function of the Melt Plug Transportation System is checked. This can lift and remove the "plug" in the reactor pit in order to make the reactor pit accessible through the core catcher. This allows the reactor pressure vessel to be inspected at maintenance intervals. At this point in time, more than 1000 km of cables had already been laid, which corresponds to 70% of the total cable length. Once the plant is completed, it is loaded with nuclear fuel for the first time and extensively tested. Once all construction defects that were discovered by the customer and the responsible supervisory authority during the final inspection have been eliminated, the nuclear power plant is accepted by the supervisory authority and the EPR is synchronized with the network.
More than 10 years can pass from the decision of an energy supply company to set up an EPR to the first kilowatt hour fed into the grid. In Great Britain, for example, the building permit alone takes at least 18 months, the subsequent earthworks almost a year, plus at least 5 years for the construction of the power plant. It takes almost another year until the network is synchronized. The construction times listed in the table on the right refer to the definition of the IAEA, so construction begins when the foundation is poured.
Over 4,000 people from different countries work on the EPR construction site, as the suppliers are spread across the globe. For example, the segments of the reactor pressure vessel from Flamanville 3 are forged by Japan Steel Works in Muroran and combined into the finished pressure vessel by Mitsubishi Heavy Industries in Kobe . Taishan's steam generators and reactor pressure vessels are manufactured on-site by Shanghai Electric Heavy Industries Group Corporation (SEC) and Dongfang Electric Corporation (DEC), respectively . Areva itself usually only manufactures an insignificant proportion of the components, usually the steam generator and pressurizer. The main contribution is of a personal nature, for example 500 employees from France, 300 from Germany and 300 from China are working on the construction site in Taishan alone. German companies are also involved as suppliers, for example Babcock Noell manufactured the steel liner for Olkiluoto 3 and Siempelkamp Nukleartechnik manufactured the core catcher.
Cost and deadline overruns during construction
In 2005, the building permit for the first EPR at the Olkiluoto nuclear power plant in Finland was granted and completion was scheduled for 2009. Completion is being delayed steadily. In September 2014, the forecast for the start of operations was postponed again and is now given as the end of 2018. The costs were originally stated at 3 billion euros for the turnkey system. These construction costs are likely to more than triple by the time of completion - even without interest and the loss of earnings due to the 9-year delay. Most recently, the expected costs were given as 8.5 billion euros when commissioning in 2015.
In 2007, construction of an EPR began in France at the Flamanville nuclear power plant . Its originally planned costs of 3.3 billion euros have risen to 9 billion euros; electricity production was announced for 2017 at the end of 2014; it was originally planned for mid-2012. According to Pierre-Franck Chevet, chairman of the ASN, the anomalies found are "very serious" and could lead to cracking. Should the prognoses be confirmed by the closer examination, there would only be the possibility of replacing the entire pressure vessel, which would mean a delay of several years and significantly increasing costs, or the abandonment of the power plant project. In addition to Flamanville, five other planned or under construction EPRs could also be affected by the problems. a. in the USA, China (Taishan) and Great Britain (Hinkley Point).
In 2008, construction of the Taishan nuclear power plant in the Chinese province of Guangdong began with two EPR units. According to Areva, these will be erected more quickly because they are counting on experience gained in Olkiluoto and Flamanville. The completion date has also been postponed several times in China.
Technical problems in construction
Of the four EPR systems under construction, two have recurring technical problems. The French Nuclear Safety Authority (ASN) reports on a pressure vessel already installed on the steel ceiling of the Flamanville plant. The carbon content in this steel ceiling is too high. In tests for resistance, the value was around 40 percent below the norm. Fine cracks can therefore form later. ASN boss Pierre-Franck Chevet said: “It is a manufacturing defect that I would describe as serious or very serious because it affects a crucial component, the boiler. The attention we pay to it is correspondingly great. "
economics
The profitability of a power plant results from the electricity production costs as well as the revenues generated on the market or the electricity exchange. The electricity generation costs result from the investment costs and dismantling costs for a power plant as well as the fixed and variable operating costs. The practical handbook for the energy industry stated electricity generation costs of 50.2 euros / MWh for a nuclear power plant operated in base load with a capacity of 1600 MW and a purchase price of 4.2 billion euros when commissioned in 2004. Since the investment costs in particular have approximately doubled since then (see below), the electricity generation costs have now risen significantly. According to current information, they are between 70 and 110 euros / MWh and thus well above the market price of electrical energy.
Investment costs
As always with electricity generation by nuclear power plants, the EPR also has comparatively high investment costs; these should be offset by low operating costs over the operating period. The investment costs of the EPR are quite high: it was originally expected to be a little over 3 billion euros per block, but H. Böck from the Atomic Institute at the Vienna University of Technology assumed in 2009 that the real price would be over 5 billion euros. Current projects in Europe stand at around EUR 8.5 to 10.5 billion as of 2015 (see table). According to Areva, plants in China can be 40% cheaper than reactors in France. There can be several reasons for this: On the one hand, the renminbi is significantly undervalued against the euro, by 48% according to the Big Mac index . Components that are manufactured in China benefit from this and cost less. On the other hand, China has considerably more construction experience in the construction of NPPs: The steel liner of Taishan 1 was completed on time and on budget, which was not possible in Olkiluoto 3 and Flamanville 3.
The following table gives an overview of the EPR projects and their costs.
| No. | Type | Location | start of building | planned end of construction | planned completion position |
estimated costs |
real construction costs |
Stei- delay |
Block power (net) |
specific investment costs |
|---|---|---|---|---|---|---|---|---|---|---|
| Billion euros | Billion euros | % | MW el | € / kW | ||||||
| 1 | FIN-EPR | Olkiluoto 3 | 08/12/2005 | 6/2009 | 2020 | 3.2 1 | 8.5 | +165.6 | 1600 | 5312 |
| 2 | Standard EPR | Flamanville 3 | 12/03/2007 | 5/2012 | 2020 | 4 2 | 10.5 | +162.5 | 1630 | 5214 |
| 3 | Standard EPR | Taishan 1 | October 28, 2009 | 12/2013 | 2018 | 3.8 3 | ? | ? | 1660 | 2289 |
| 4th | Standard EPR | Taishan 2 | 04/15/2010 | 11/2014 | 2018 | 3.8 3 | ? | ? | 1660 | 2289 |
| 5 | UK EPR | Hinkley Point C1 | 2019 | 2023 | 2025 | £ 8 billion (approx. € 9.5 billion) | - | - | 1630 | approx. 5800 |
| 6th | UK EPR | Hinkley Point C2 | 2017 | 2023 | 2025 | £ 8 billion (approx. € 9.5 billion) | - | - | 1630 | approx. 5800 |
| 7th | UK EPR | Sizewell C1 | not yet determined | ? | ? | ? | 1630 | ? | ||
| 8th | UK EPR | Sizewell C2 | not yet determined | ? | ? | ? | 1630 | ? |
Generation costs
In principle, the generation costs of a power plant are always calculated according to the same scheme: Based on the specific investment costs, the utilization and the design life of the power plant, the costs incurred during the operation of the plant are added.
The generation costs of a nuclear power plant are composed of operating costs, maintenance costs, fuel and disposal costs. In some studies, such as in “Comparison of electricity generation costs” by the Technical University of Lappeenranta (LUT) by Risto Tarjanne and Aija Kivistö from 2008, the disposal costs are also included in the operating and maintenance costs. These are listed separately below:
- Capital costs : A significant cost item when operating a nuclear power plant are the capital costs for the construction of the power plant, which are usually covered by both equity and debt. When taking out outside capital, three factors are decisive: the amount of the loan, its interest rate and its duration. All studies are based on annuity loans , which increases the generation costs by a constant amount (in this case by 2.0 euro cents / kWh) in the depreciation period. For example, the LUT study cited above assumes a 40-year depreciation period, 5% real interest rate, and 100% debt financing. The study also notes that the full financing of a power plant with borrowed capital is a conservative assumption , as is the interest rate of 5%, which was about 2% higher than the market rate at the time. Since EVUs usually have an A rating, the interest rates are relatively low. In the study, the investment costs were set at 2750 € / kW, the output at 1500 MW, which resulted in investment costs for a power plant unit of 4.125 billion euros. However, these figures are now out of date. As of 2013, the investment costs for the reactors under construction in Europe in Olkiluoto and Flamanville are around 5200-5300 € / kW and thus around twice as high as assumed in the 2008 study. The planned power plants at the Hinkley Point location are around € 5800 / kW (see table above). The capital costs increase accordingly in the same ratio to 3.8–4.0 euro cents / kWh (Olkiluoto and Flamanville) and 4.3 euro cents / kWh (Hinkley Point).
- Operating costs : These costs are incurred when operating the system, including a. for staff, inspections and power requirements during idle times. Due to its higher unit output compared to previous power plants, the EPR may achieve savings effects due to the degression effect - more electricity production with fewer personnel and fewer components per kilowatt hour. The United States Department of Energy (DoE) reports generation costs of 1 ct / kWh for existing (i.e., depreciated) US nuclear power plants. However, this figure also includes the American disposal costs.
- Maintenance costs : Maintenance costs arise when components have to be replaced or the system's performance is increased with a new steam generator or a new turbo generator. Compared to the previous models, the EPR has 16% fewer pump and turbine parts, 23% fewer components in the heat exchangers, 30% fewer tanks and 26% fewer valves. Areva therefore states a saving of 35% in maintenance costs. The LUT study indicates operating and maintenance costs at 1 ct / kWh, which also includes disposal costs. The Nuclear Energy Institute (NEI) gives 1.49 $ cents / kWh (1.1 € cents / kWh) for operating and maintenance costs for the existing power reactors in the USA.
- Fuel costs : The 235 U share in the EPR nuclear fuel must be between 1.9 and 4.9%. The price for uranium (V, VI) oxide and uranium separation work can be viewed on the exchange on a daily basis. The use of MOX fuel elements causes significantly higher fuel costs, which depend on the origin of the plutonium.
In the study by the Technical University of Lappeenranta, for example, generation costs are calculated in the depreciation period of 3.5 ct / kWh, since depreciation is stretched over 25 years and the investment costs are lower.
Areva itself indicates that the production costs by at least 10% lower than in existing nuclear power plants, with 1,500 MW e performance. These calculations do not include the costs for the disposal, reprocessing and final storage of radioactive fission products.
Data tables
- Note: The information relates to the first EPR in the Olkiluoto NPP
| Technical specifications: | |
|---|---|
| Thermal performance | 4300 MW th |
| Generator power | 1720 MW |
| Electric power (net) | 1600 MW |
| Efficiency (net) | 37% |
| Own electrical consumption | 120 MW |
| Projected operating time | 60 years |
| Total volume of the power plant | 1,000,000 m³ |
| Nuclear reactor: | |
| Nuclear fuel | UO 2 |
| Number of fuel elements | 241 |
| Fuel rods per fuel assembly | 265 |
| Length of fuel assemblies | 4.8 m |
| Active height of the core | 4.2 m |
| Diameter of the core | 3.77 m |
| Fuel mass | about 128 tons of uranium |
| enrichment | 1.9-4.9% fissile material |
| Burn-off | 45 GWd / t |
| Gap fraction in plutonium mixed oxide fuel | 50% 239 Pu; 27% 235 U; 14% 241 Pu; 9% 238 U |
| Absorber bundle | 89 |
| Limit neutron fluence (> 1 MeV ) for pressure vessel | about 10 19 n / cm² |
| Average heating rate per fuel rod | 156.1 W / cm |
| Energy density of the core | about 91.7 MW / m³ |
| Reactor inlet temperature | 296 ° C |
| Reactor outlet temperature | 328 ° C |
| Centrifugal pumps: | |
| number | 4th |
| Mass flow per pump | 23,135 kg / s |
| Pressure in the circuit | 155 bar |
| Maximum head | 102 m, ±% 5 |
| Rotation speed | 1465 rpm |
| Power requirement per pump | 9 MW |
| Pressurizer: | |
| number | 1 |
| Design pressure | 176 bar |
| Design temperature | 362 ° C |
| Empty mass | 150 t |
| Relief valves | 3 × 300 t / h |
| Safety valve (rupture disc) | 1 × 900 t / h |
| Steam generator: | |
| number | 4th |
| Heat exchanger area per steam generator | 7960 m² |
| Number of tubes per steam generator | 5980 |
| Total mass | 520 t |
| Feed water temperature | 230 ° C |
| Superheated steam temperature | 293 ° C |
| Vapor pressure | 78 bar |
| Steam mass flow | 2443 kg / s |
| Turbine: | |
| number | 1 |
| Steam pressure high pressure turbine | 75.5 bar |
| Number of high pressure turbines | 1 |
| Number of low pressure turbines | 3 |
| rotational speed | 1500 rpm |
| Overall diameter | 6.72 m |
| Length of the turbo set | 68 m |
| Turbine exit surface | 180 m² |
| Generator: | |
| number | 1 |
| Nominal power | 1992 MVA |
| Effective performance | 1793 MW el |
| Magnetizing current | 9471 A |
| Power factor | 0.9 |
| Cooling gas | hydrogen |
| Capacitors: | |
| number | 6th |
| Cooling surface | 110,000 m² |
| Cooling water volume flow | 57 m³ / s |
| Condenser pressure | 24.7 mbar |
| Feed water: | |
| Feed water pumps | 4th |
| Feed water preheater | 7 levels |
| Security technology: | |
| Containment volume | 80,000 m³ |
| Design pressure | 5.3-5.5 bar |
| Number of security containers | 2 |
| Emergency cooling systems | 4 × 100% |
| Emergency feed in steam generator | 4 × 50% |
| Maximum ground acceleration | 0.25 g |
Web links
- EPR ™ reactor: the very high power reactor Manufacturer's website for the Areva EPR
- Public Version of UK EPR ™ GDA Submission By clicking on the tabs and expanding the tree structure, all technical documents of the UK EPR that are necessary for approval can be viewed
Individual evidence
- ↑ a b c Andrew Teller, Areva: The EPR ™ Reactor: Evolution to Gen III + based on proven technology ( Memento from January 31, 2012 in the Internet Archive ) (PDF; 746 kB)
- ↑ IAEA Studsvik Report: ADVANCED NUCLEAR REACTOR TYPES AND TECHNOLOGIES (PDF; 7.5 MB)
- ↑ LRST Aachen: EU-CoMaS ( Memento of the original from December 4, 2017 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b Siempelkamp Nukleartechnik: Core catcher components for EPR ( memento of the original from March 21, 2012 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b Society for Plant and Reactor Safety (GRS) - Simulation of Melt Spreading in Consideration of Phase Transitions ( Memento of the original from January 19, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 1.2 MB)
- ↑ a b CEA: The VULCANO Spreading Program (PDF; 341 kB)
- ↑ Ex-Vessel Core Melt Stabilization Research (ECOSTAR) ( page no longer available , search in web archives ) Info: The link was automatically marked as defective. Please check the link according to the instructions and then remove this notice.
- ↑ Handelsblatt: Outlawed at home, coveted internationally
- ↑ EDF: EDF: The first of two EPR reactors at China's Taishan nuclear power plant enters into commercial operation. December 14, 2018, accessed December 14, 2018 .
- ↑ world nuclear news: World's second EPR starts operations on September 9, 2019
- ↑ 36-hour Hinkley Point concrete pour releases £ 70m | Construction inquirer. Retrieved December 10, 2018 .
- ↑ a b c Differences in the staggered security concept: Comparison of Fukushima Daiichi with German systems ( memento of the original from January 8, 2012 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b c KIT: Chapter 3: Advanced Light Water Reactors ( Memento of the original from January 18, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 4.4 MB).
- ↑ a b JAITAPUR EPR REACTOR ( Memento of the original from November 14, 2011 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 48 kB)
- ↑ Gen IV Forum - Basis for the Safety Approach for Design & Assessment of Generation IV Nuclear Systems: “In parallel, versus the Gen II systems, the quantitative safety objectives applicable to the reactors of the third generation (eg AP1000 and EPR) are very ambitious and guarantee an improved level of protection reducing the level of risk in a demonstrable way. The RSWG believes that this achieved level is excellent and can be kept as a reference for future reactors. ” ( Memento of the original from January 8, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 1.3 MB)
- ↑ AREVA: Tests of the passive safety systems of the SWR 1000 on the newly built INKA test stand in Karlstein ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ - ( Memento of the original from May 15, 2015 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ Areva / Mitsubishi ATMEA1 (PDF; 1.1 MB)
- ↑ ATMEA1 reactor ( memento of the original from September 23, 2015 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b Visiatome - Nucléaire du Futur / Enjeux et Technologies / La voie alternative du thorium ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link has been inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 1.3 MB)
- ↑ a b Center national de la recherche scientifique (CNRS) - Parcs symbiotiques RNR-EPR cycle U et Th (PDF; 5.6 MB)
- ↑ Experts call for retrofitting of French nuclear power plants . In: Süddeutsche Zeitung , November 17, 2011, accessed on November 18, 2011
- ↑ a b c d e f g h i j k l m n o p q r s t TVO (PDF; 7.4 MB)
- ↑ a b c d e f g h i j US EPR ( Memento of the original from November 23, 2011 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ [1]
- ↑ WNA - Plutonium / Areva's EPR design is capable of running a full core load of MOX.
- ↑ Areva - EPR reactor: the very high power reactor / An EPR power plant can operate with uranium enriched up to 5%, reprocessed uranium or MOX fuel (in variable proportions according to customer needs and up to 100%)
- ↑ Forschungszentrum Jülich: Optimized conversion of plutonium and americium in pressurized water reactors ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 11.8 MB) The graphic is shown on page 43/185. At around 50 MWd / kg for a convoy system, around 15% Pu-241 and 50% Pu-239 result
- ^ Thorium Power, Ltd. - Rodman & Renshaw Presentation September 2009 (The company changed its name to Lightbridge Corporation on September 29, 2009.)
- ↑ Advanced reactors with innovative fuels: second workshop proceedings - Thorium Fuel in LWR: An option for an effective reduction of plutonium stockpiles
- ↑ a b c Vienna University of Technology / Austria (PDF; 3.9 MB)
- ↑ Siemens AG - Raising steam on an unprecedented scale (PDF; 474 kB)
- ↑ Alstom - ARABELLE ™ Steam Turbine for Nuclear Power Plants ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ Alstom - Technical Project Datasheet ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b MAIN FEATURES OF THE CORE MELT STABILIZATION SYSTEM OF THE EUROPEAN PRESSURIZED WATER REACTOR (EPR) ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b c UK-EPR: CHAPTER 6: CONTAINMENT AND SAFEGUARD SYSTEMS (PDF; 2.0 MB)
- ↑ a b EDF - EPR Technology Fact Sheet ( Memento of the original dated February 8, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b c UK-EPR: FUNDAMENTAL SAFETY OVERVIEW: "Redundant 100% capacity safety systems (one per safeguard building) arranged in four trains are strictly separated into four divisions. This divisional separation is provided for electrical and mechanical safety systems. The four divisions of safety systems are consistent with an N + 2 safety concept. With four divisions, one division can be out-of-service for maintenance and one division can fail to operate, while the remaining two divisions are available to perform the necessary safety functions even if one is ineffective due to the initiating event. " ( Memento des Originals from January 31, 2012 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ US EPR Standard Design Certification, AREVA NP Inc. (PDF; 77 kB)
- ↑ a b c Areva - EPR ™: The safest GEN III + reactor ( memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 917 kB)
- ↑ THE VALIDITY OF SAFETY GOALS (PDF; 65 kB)
- ↑ Nomination of Land adjacent to Oldbury Nuclear Power Station - Further information on flood risk, storm surge and tsunami ( Memento of the original dated January 11, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ STUK - PRELIMINARY SAFETY ASSESSMENT ON OLKILUOTO 4 NUCLEAR POWER PLANT PROJECT ( Memento of the original from January 18, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 1.2 MB)
- ↑ US EPR Mechanical Systems ( Memento of the original from January 11, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ UK-EPR: FUNDAMENTAL SAFETY OVERVIEW VOLUME 2: DESIGN AND SAFETY (PDF; 225 kB)
- ↑ UK-EPR: PCSR - Sub-chapter 6.6 - Emergency Feedwater System (ASG) (EFWS) (PDF; 306 kB)
- ↑ PCSR - Sub-chapter 15.3 - PSA of accidents in the spent fuel pool ( Memento of the original from November 27, 2011 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ a b Tim Stack / Areva: EPR Severe Accident Design Features ( Memento of the original from January 11, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ Application for a desicion-in-principle concerning the construction of an nuclear power plant unit - Olkiluoto 4 / Ensuring fuel integrity (PDF; 2.7 MB)
- ↑ SMiRT-15 EPR Accident Scenarios and Provisions ( page no longer available , search in web archives ) Info: The link was automatically marked as defective. Please check the link according to the instructions and then remove this notice.
- ↑ UK-EPR: PCSR - Sub-chapter 15.3 - PSA of accidents in the spent fuel pool ( Memento of the original dated November 27, 2011 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ AP1000®: A Simple, Safe, and Innovative Design that leads to Reduction in Safety Risk ( Memento of the original of November 22, 2011 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 1.9 MB)
- ↑ Department of Energy - The Westinghouse AP1000 Advanced Nuclear Plant ( Memento of the original from December 18, 2011 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ Introduction and General Description of Plant - AP1000 Design Control Document (PDF; 451 kB)
- ↑ GE-Hitachi - ESBWR Overview ( Memento of the original from January 11, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ The TELEPERM XS Qualified Display System. Overview, Concept and Applications (PDF; 87 kB)
- ↑ Joint Regulatory Position Statement on the EPR Pressurized Water Reactor ( page no longer available , search in web archives ) Info: The link was automatically marked as defective. Please check the link according to the instructions and then remove this notice.
- ↑ world nuclear news: EPR system modifications satisfy UK regulator
- ↑ ARIS ( page no longer available , search in web archives ) Info: The link was automatically marked as defective. Please check the link according to the instructions and then remove this notice.
- ^ World Nuclear Association - Load following with PWR nuclear plants
- ↑ spiegel.de from February 24, 2009: The French build nuclear power plants in Italy
- ↑ Financial Times Deutschland, June 13, 2011: Serious defeat for Berlusconi in the atomic vote ( Memento from August 16, 2011 in the Internet Archive )
- ↑ FAZ, June 13, 2011: Italy votes against nuclear power - and against Berlusconi
- ^ La Repubblica: Speciale elezioni 2011 , accessed on June 16, 2011
- ^ A b Second Chinese EPR achieves criticality - World Nuclear News. Accessed May 31, 2019 .
- ^ The second EPR reactor at China's Taishan nuclear power plant about to enter into commercial operation. September 6, 2019, accessed October 5, 2019 (en-en).
- ↑ Areva no longer builds nuclear power plants - network IT
- ↑ Olkiluoto 3 reactor delayed yet again, now 12 years behind schedule. Retrieved January 10, 2020 .
- ^ Word Nuclear Association - Nuclear Power in the United Kingdom
- ↑ Exchange rate from.
- ^ Britain, EDF strike deal on nuclear project . In: Global Post , October 17, 2013. Retrieved October 18, 2013.
- ↑ England's new nuclear power is more expensive than solar power . In: Wirtschaftswoche , October 21, 2013. Retrieved October 23, 2013.
- ↑ https://www.edfenergy.com/energy/nuclear-new-build-projects/hinkley-point-c/about
- ↑ http://sizewell.edfenergyconsultation.info/
- ↑ http://www.world-nuclear-news.org/NN-Calvert-Cliffs-3-COL-withdrawn-2107157.html
- ^ NRC - Location of Projected New Nuclear Power Reactors
- ^ Message from the Nuclear Forum
- ↑ NRC: US EPR FINAL SAFETY ANALYSIS REPORT (PDF; 299 kB)
- ↑ a b Areva: EPR design and status of the current EPR projects ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 4.8 MB)
- ^ WNA: Reactor dome installed on Chinese EPR
- ↑ Areva Field Report # 3 (PDF; 547 kB)
- ↑ Areva Field Report # 4 (PDF; 815 kB)
- ↑ Areva Field Report # 5 (PDF; 793 kB)
- ↑ Areva New Build Field Report # 7 (PDF; 808 kB)
- ↑ Areva: Finland - Olkiluoto 3
- ↑ Areva: Newbuid EPR Reactors ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ^ WNA: Heavy Manufacturing of Power Plants
- ^ Areva: Taishan 1 & 2 - China: Supply chain
- ↑ BABCOCK NOELL: EPR Olkiluoto, Finland
- ↑ Olkiluoto 3 to be ready in 2018 . In: Helsinki Times , October 9, 2014. Retrieved October 9, 2014.
- ↑ The grave of billions . taz. October 9, 2014. Retrieved October 9, 2014.
- ↑ a b Serious defects in the Flamanville breakdown nuclear power plant . In: Handelsblatt , April 17, 2015. Retrieved April 18, 2015.
- ↑ a b c Share of the world's largest nuclear company crashes . In: Die Tageszeitung , November 20, 2014. Retrieved April 18, 2015.
- ↑ areva.com: Taishan 1 & 2 - Key Milestones (last updated December 2012), accessed October 28, 2014.
- ↑ http://www.taz.de/Sicherheitsprobleme-in-Atomanlagen/!158566/
- ↑ Panos Konstantin, Practical Guide to Energy Economics. Energy conversion, transport and procurement in the liberalized market. Berlin - Heidelberg 2009, p. 302.
- ↑ The French have doubts about nuclear power . In: Die Zeit , December 6, 2012. Retrieved January 2, 2014.
- ^ E.ON and RWE overturn nuclear power plant plans in Great Britain . In: Reuters , March 29, 2012. Retrieved January 2, 2014.
- ^ Britain, EDF strike deal on nuclear project . In: Global Post , October 17, 2013. Retrieved January 2, 2014.
- ↑ http://www.taz.de/Probleme-beim-Reaktor-Bau-in-Frankreich/!5229699/
- ↑ Bloomberg: China Builds Nuclear Reactor for 40% Less Than Cost in France, Areva Says
- ↑ Bloomberg: French Nuclear Watchdog Says EDF Has Problems With Flamanville EPR Liner
- ↑ Delays cause new problems at Siemens . In: Handelsblatt , February 11, 2013. Accessed February 11, 2013.
- ↑ a b Taishan EPR Nuclear Reactor Project Delayed. powermag, February 23, 2017, accessed on June 21, 2017 .
- ^ Hinkley Point C gets go-ahead for construction . March 28, 2017. Retrieved June 21, 2017.
- ^ A b Hinkley Pont C: Strike action threat over bonuses averted . June 7, 2017. Retrieved June 17, 2017.
- ↑ a b Both blocks together are expected to cost around 19 billion euros. Criticism of British nuclear plans . In: Frankfurter Allgemeine Zeitung , October 22, 2013. Retrieved October 23, 2013.
- ↑ WNN: Areva expects losses after extra Olkiluoto provision
- ^ WNA: Nuclear Power in France
- ↑ TPR Engineering: NUCLEAR POWER PLANT EPR Flamanville - France ( Memento of the original of October 3, 2013 in the Internet Archive ) Info: The archive link was automatically inserted and not yet checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ Areva: The Taishan 1 & 2 project ( Memento of the original from January 31, 2012 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 2.9 MB)
- ↑ WNA: Nuclear Power in China, Taishan 1 & 2 EPR
- ↑ a b c d e Comparison of electricity generation costs ( Memento of the original from October 14, 2011 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice. (PDF; 357 kB)
- ^ EIA: Average Power Plant Operating Expenses for Major US Investor-Owned Electric Utilities
- ^ NEI: Costs: Fuel, Operation and Waste Disposal
