The Dukovany Nuclear Power Station
- The Dukovany Nuclear Power Station
- Engineering and technology
- Safety of The Dukovany Nuclear Power Station
- Basic definitions
- Nuclear fuel
- Professional training of the personnel
The Dukovany Nuclear Power Station
The Dukovany Nuclear Power Plant is situated approximately 30 km southeast of Třebíč in a triangle formed by the municipalities of Dukovany, Slavětice and Rouchovany. Four pressurised-water reactors of Type VVER 440 – Model V 213. Each of these reactors has the heat capacity of 1,375 MW and electric capacity of 510 MW.
Its construction was started in 1974. A change to the project caused postponement of the full start-up of the construction until 1978. The first reactor unit was put into operation on 4 May 1985 and the last (the fourth) unit on 20 July 1987. The Dalešice waterworks with a pumped-storage hydroelectric power plant of 450 MW capacity was built in the vicinity of the power plant on the Jihlava River. Its equalising reservoir serves as a water resource for the nuclear power plant.
The Dukovany Nuclear Power Plant is intended for a base-load operation mode. It annually supplies approximately 13 TWhr of electric energy to the national power network. Particular attention is paid to the safety of its operation which is supervised on on-going bases by the State Office on Nuclear Safety and relevant international organisations.
The reactors are fuelled by uranium dioxide UO2. Fuel is placed in the reactor in 312 fuel assemblies. Each assembly consists of 126 fuel rods with a hermetically sealed fuel. In addition, the reactor contains 37 control rod assemblies with the fuel part. Improved nuclear fuel parameters enabled a smooth transition in 1997 from a three- to four-year fuel cycle, and since 2003 five-year cycle have been successively started.
The heat in the reactor core generated by the controlled fission of the uranium-235 nuclei is removed from the reactor by cooling demineralised water that also serves as a neutron moderator. In addition, an admixture of boric acid (max. 12 g/l of water) also contributes to the reactor output regulation.
The cooling water, sealed under high pressure in the primary reactor circuit, is circulated by means of six independent cooling loops with pumps and steam generators. The water of the closed primary circuit water passes its heat into the secondary circuit in the steam generators. The secondary circuit is also closed and filled with demineralised water. The secondary circuit water is converted into steam in the steam generators to drive the turbines. Each reactor forms a set with two three-casing turbines, each with one high-pressure and two low-pressure sections running at 3,000 revolutions per minute. There are eight such turbines in the power plant. Coupled to each turbine is a 220 MW power generator (double-pole asynchronous alternator generating voltage of 15.75 kV).
Downstream from the turbines the steam of the secondary circuit condenses back into water in the large condensers cooled by water from the tertiary cooling circuit. This circuit is than led out into cooling towers where the tertiary circuit water is recooled with natural air flow.
The power plant modernisation will successively be carried out to the end of its planned service life. At present, however, in view of the high quality of its main structural components, an extension of its service life by ten years, i.e. to a total of forty years of operation is being considered.
A storage site for low and medium radioactive waste is situated in the area of the power plant. In 1995, a dry storage facility for the spent nuclear fuel was completed in the area of the power plant, and after its trial operation an on-going operation was started, its capacity is sixty CASTOR 440/84 containers. The capacity of one such container (10 tons of spent nuclear fuel) corresponds to the amount of waste produced by one VVER 440 reactor in almost one year. Nevertheless, this storage facility will not be able to take all the spent fuel produced by the power plant during its entire service life. Thus a new facility should be put into operation by no later than 2005 allowing spent nuclear fuel storage for a period not exceeding 50 years. If, in the meantime, the spent fuel is not utilised as a valuable source of raw materials (at present it is not feasible to recycle spent nuclear fuel both from the financial and technological point of view), or the so-called transmutation of the spent fuel is not available on a large scale, the spent nuclear fuel will be moved to a permanent deep geological repository.
|Generation Unit – Dukovany|
|Installed capacity||4 x 510 MW|
|Year of commissioning||1985 – 1988|
|Type of reactor||VVER 440|
Engineering and technology
The Dukovany Nuclear Power Plant is the first nuclear power plant operated in the Czech Republic, and belongs to the largest, highly reliable and economically profitable power resources of ČEZ, a. s. The annual production of electric energy is approximately 13 TWhr, which represents about 20% of the total consumption of electricity in the Czech Republic. In comparison with other important producers of electricity, the Dukovany Nuclear Power Plant produces electricity with the lowest specific costs.
Four pressurised-water reactors (PWR) are installed in the Dukovany Nuclear Power Plant. The design-mark of these reactors is VVER 440/213. The abbreviation VVER (in Czech) means Water-cooled, Water-moderated Energy Reactor. Each of these reactors has the heat capacity of 1,375 MW and electric capacity of 510 MW.
The layout of the power plant includes two main production units. Each of them contains two reactors and all directly linked equipment, including the machine hall with turbines and generators.
Aside from the four reactor units, there are also another two nuclear facilities located in the Dukovany Nuclear Power Plant area:
Spent nuclear fuel storage facility where spent fuel is safely stored in the transportable CASTOR 440/84 storage containers.
Low and medium radioactive waste repository under state ownership, 7% of the area is occupied.
The first reactor unit of the Dukovany Nuclear Power Plant was commissioned in 1985, the second and third units in 1986, and the fourth unit in 1987.
The following organisations participated in the design, production of the equipment and construction of the power plant:
Project base documents: LOTEP Company (former Soviet Union)
Execution project: Energoprojekt Praha
General contractor of the construction: Průmyslové stavby Brno
General contractor of the technology: Škoda Praha
|Design, production and delivery of decisive equipment:|
Dukovany Nuclear Power Plant Safety
The safety of the nuclear power plant is achieved by the design safety and the power plant’s operational culture level, which includes qualified personnel, quality documentation, use of operating experience, technical control, protection against radiation, fire safety, etc.
Safe designNuclear power plant equipped with VVER 440/213 reactors have some important design advantages. For example, the pressure vessel of the reactor and the primary circuit piping include a very small content of cobalt. This results in a lower activation of material and a lower irradiation of personnel. Strong feedback during output operation of the reactor ensures the reactor stability without xenon oscillations.
Safe operation of the power plant is ensured by six shifts of the same standard. The control shift personnel – operators in the unit control room – consists of seven shifts. The seventh shift is formed due to the high demands that are made on periodic training of the control personnel. The highest shift chief of the entire nuclear power plant is the shift engineer. Each of the four reactor units is controlled from the independent unit control room. The attendance of the unit control room includes the reactor unit chief, primary part operator and the secondary part operator.
Inspections of technical conditions in the power plant are conducted by the power plant employees as well as independent supervision bodies and inspection institutions. Technical inspections are performed regularly by trained workers in accordance with the pre-approved procedures. The latest technologies are used during technical inspections. The most stringent inspections are focused on the facilities that are important in view of nuclear safety. The nuclear safety of the nuclear power plant is not determined by its actual condition at the time of final inspection before its commissioning.
The requirements of our supervising bodies and recommendations from the Atomic Energy International Agency develop permanent pressure on continuously increasing nuclear safety in power plants until termination of their operation. An example of the already realised actions leading to increased nuclear safety includes fire-suppression spraying of cables, installation of the new public warning systems in case of emergency, formation of the crisis centre, elaboration of new regulations for liquidation of crash conditions or replacement of the electric power supply of the 1st category equipment.
Safety systems The basic precondition of power plant safety is the continuous removal of heat generated in the reactor core. The safety systems of the Dukovany Nuclear Power Plant consist of high- and low-pressure emergency pumps, sprinkler system pumps, reservoirs with boric acid solution, heat exchangers, pressurised-water containers, pipelines, fittings, barbotage (condensing) troughs, barbotage towers and gas tanks. In case of accidents connected with the leakage of cooling water from the primary circuit, the pressure of the primary circuit cooling water would be reduced while the steam pressure in the hermetic boxes would simultaneously be increased. Depending on the kind of accident and leakage extent, the security systems would pump cooling water under and over the reactor core and sprinkle the hermetic boxes. In case of disruption of the main circulation piping, the pressure of steam generated in hermetic boxes is increased to such a level that part of the steam would flow into the barbotage (condensing) troughs where it would condense. Despite the fact that the occurrence of such an event is highly improbable, these facilities are doubled or tripled, and dimensioned in such an extent that even in case of an accident of this kind, the leakage of radioactive substances into the environment would be reduced to a minimum.
Harmonisation Programme The actual programme for increasing the safety of the Dukovany Nuclear Power Plant is included in the newly outlined Harmonisation Programme. This programme is not only focused on questions connected with the power plant project change and replacement of some equipment, but also covers other fields that can affect the safety of the power plant. The highest importance or contribution to increasing safety need not be, as generally assumed, improvement of the facilities. A higher level of safety can be achieved by improving the safety culture as well. The Dukovany Nuclear Power Plant’s target is to achieve, by implementation of the Harmonisation Programme, a reduction of the reactor core damage probability coefficient from the present value of 1.7*10-5 to the value of 7.7*10-6 in the year 2010 (this value means that an event causing fuel damage in the reactor core can occur probably once in a period of 130,000 years). This value has been recommended for newly built power plants by the Atomic Energy International Agency.
It includes a system of facilities that enables the acquisition of heat energy from nuclear fuel through controlled chain fission reaction, its continuous removal by means of coolant, and its conversion to a form of heat energy usable in the steam turbine.
Basic facilities of this circuit include:
- Steam generators
- Main circulating pumps
- Circulation piping of the primary circuit
- Volume compensator
As already mentioned above, it is a resource of heat for heating the pressurised water released during the controlled chain fission reaction in the nuclear fuel. Heat released from the reactor core is removed by the forced circulation of coolant that is carried out by the main circulation pumps.
The nuclear reactor is a technical facility (containing nuclear fuel, coolant, moderator, structural materials and control systems) that is intended for maintaining the controlled chain fission reaction and enabling a smooth removal of heat energy released during the fission process. It consists of a steel pressure vessel equipped with a removable cover. Located inside the reactor is a core that contains the nuclear fuel and regulation equipment for control and monitoring the fission reaction.
Circulating (main circulating) pump
The main circulating pump provides circulation of coolant in the primary circuit in a quantity corresponding to the heat capacity of the reactor. In the viewpoint of the design, it is a vertical-shaft centrifugal stuffing-box pump driven by an asynchronous electric motor.
Despite the fact that the volumetric-thermal expansivity coefficient of water is relatively small, the growth of volume due to temperature influence as the volume of the primary circuit coolant reaches a few hundreds of cubic metres has to be taken into consideration. In case that the growth of coolant volume is not compensated in some way, the volume growth of water causes such a high mechanical stress of the primary circuit facilities that it would result in the disruption and release of radioactive coolant into the primary circuit area. The volume compensator is a vertical steel pressure vessel with a size comparable to the reactor pressure vessel, connected by piping to a hot branch of one of the primary circuit loops. Apart from compensation of volumetric-thermal changes of the coolant, the volume compensator also serves for regulation of the primary coolant pressure by means of the built-in electric heaters or showers. The volume compensator is equipped with safety valves against exceeding the permissible pressure value in the primary circuit.
The horizontal pressure evaporator heat exchanger is where the primary circuit water (flowing in pressure pipes in the steam generator) transfers its heat to the secondary circuit water. As the temperature of water in the primary circuit is higher than temperature of the boiling point of water in the secondary circuit (pressure of water in the primary circuit is in fact more than doubled in comparison with the pressure of water or steam in the secondary circuit), it effects a more intensive generation of steam in the generator that is taken through steam piping to the turbine.
Primary circuit piping
Stainless steel piping of 500 mm diameter and a wall thickness of 32 mm interconnects the reactor, steam generator and circulating pumps. To reduce heat loss but simultaneously to enable control of its condition, this piping is equipped with a removable heat insulation. That part of the piping, located between the reactor and the steam generator in which the heated water flows from the reactor to the steam generator, is called the hot branch, and the outstanding part of the piping that removes water from the steam generator through the circulating pump into the reactor is called the cold branch of the primary circuit.
The secondary circuit in the power plant includes a system of facilities that enable the conversion of heat energy from steam to the mechanical energy from the steam turbine rotor.
Basic facilities of this circuit include:
- Turbine and generator
- Condensate and feeding pumps
- Regenerative heaters
Turbine and generator
Rotary heat motor in which the internal energy of steam is converted into a rotary mechanical energy in the turbine rotor. Upon impulse, the turbines pressure drop of steam changes in the stationary blades of the stator to a kinetic energy of steam that is transferred through the moving blades to the rotor. The turbine rotor is linked with the generator rotor where kinetic energy of the rotor is transformed to electric energy.
Heat exchanger where steam condenses after expansion in the turbine and cooling by cooling water. It is placed in the immediate vicinity of the bottom part of the low-pressure section of the turbine. Steam leaving the turbine passes between pipes in which cooling water is flowing, and condenses on the surface of pipes. The condensed steam (the condensate) is transported by the condensate pumps through the condensate treatment section, regenerative exchangers and degasification section to the steam generator.
Low-pressure and high-pressure regenerative heaters
Heat exchangers where steam from non-regulated regenerative take-offs of the turbine transfers its condensation heat to the condensate or feeding water of the steam generator. In low-pressure regenerative exchangers the condensate is successively heated to boiling point so that the gases dissolved in it can be removed in the degasification tank. In high-pressure regenerative heaters the feeding water, after removal of gases in the degasification tanks, is heated to temperatures close to boiling point in the steam generator.
Condensate and feeding pumps
Condensate pumps are intended for pumping the condensate from the turbine condensers through low-pressure regenerative heaters into the degasification tank. Feeding pumps transport the degasified feeding water from the degasification tank through high-pressure regenerative heaters to the steam generator, and simultaneously increase the pressure of the degasified feeding water to pressure of the generated steam.
A task of the tertiary circuit is to create the highest vacuum, usable by the turbine, in the condenser to achieve the highest possible efficiency of the turbine. The lower the temperature of the cooling water in the tertiary circuit, the higher the vacuum created in the condenser.
Basic equipment of this circuit includes:
- Cooling towers
- Circulating pumps
- Cooling water piping and channels
Power plants built in the vicinity of the sea or large rivers are not equipped with cooling towers as the condenser can be cooled with the through-flow water, without fears of a negative impact of the heated water to the water ecosystem.
They represent the dominating feature of power plants but at the same time the cooling towers are subtle constructions made from reinforced concrete in the shape of a hyperboloid of revolution that serve for providing the sufficient draught of cooling air needed for cooling the cooling water, and for mounting built-in structures that ensure the sprinkling of cooling water aimed at the better efficiency of its cooling. A part of the cooling water is evaporated. Latent heat needed for evaporation is the main reason of temperature drop of cooling water. In the bottom section of the cooling tower is a round pool in which the cooled water is collected and transported back by means of the cooling water pumps to the turbine condenser.
Centrifugal pumps serving for circulation of cooling water between condensers of turbines and cooling towers.
Cooling water piping and channels
Cooling water flow can be compared with a flow of water in a river. It is the piping of the largest diameter in the entire power plant.
One of the principle requirements for radiation safety in nuclear power plants is to prevent an uncontrolled leakage of radioactive substances into the environment. Therefore, radioactive substances are separated from the life environment by a few barriers.
The first barrier is the actual fixation of the radioactive substances in the fuel cells. The second barrier is formed by the hermetically sealed fuel rods in which the cells are sealed. The third barrier represents a tightly sealed primary circuit, and the fourth barrier is the containment.
Fresh nuclear fuel
Fresh fuel includes approximately 4% of uranium-235 isotope. As natural uranium contains only 0.7% of uranium-235, prior to the production of the fuel cells the so-called uranium enrichment must be carried out. In the course of operation the contents of uranium-235 is reduced by the fission process. The Dukovany Nuclear Power Plant used fuel that has been designed for three-years use in the reactor (the so-called three-year fuel campaign). Presently, there is also fuel designed for the four-year fuel campaign.
Spent nuclear fuel
Fuel assemblies containing spent nuclear fuel removed from the reactor look the same as the assemblies containing the fresh fuel. They are clean and undamaged. However, there is a significant difference in radioactivity of substances contained in the fuel assembly. In the course of operation radioactivity increases from almost zero with a successive increase of the quantity of fission products in nuclear fuel. It is caused mostly by the fact that fission of one uranium-235 atom results in the creation of two unstable atoms of various elements that continue in conversion. Therefore, even after removing the nuclear fuel from the reactor, the nuclear transmutation and release of gamma radiation, neutrons and heat that must be removed continues.
What is the function of nuclear fuel? |Nuclear fuel contains a minor quantity of uranium-235 isotope. If the uranium-235 atom meets a slow neutron, it is fissured into two atoms of the lighter elements, namely into two or three fast neutrons. Practically at the same time, energy in the form of gamma radiation and heat is released. To increase or reduce the reactor output use a lifting and lowering control rod assembly.
Control rod assemblies
The control rod assembly height is approximately doubled in comparison with the ordinary fuel assembly that form the reactor core located in the very centre of the reactor vessel. The lower half of the control rod assembly is the same as the fuel assembly; the upper half is made from materials that absorb neutrons. If the control rod assembly is lowered into its bottom position, the part that absorbs neutrons is placed in the reactor core. When lifting up the control rod assembly, its lower half containing the nuclear fuel is gradually inserted into the reactor core, and results in increasing the reactor output.
Designed fuel charge
The designed fuel charge assumes use of the nuclear fuel in the so-called three-year cycle, which means that every fuel assembly operates in the reactor for the period of three years and then is placed into the spent fuel pool and substituted by a fresh fuel assembly. Every year approximately one third of the fuel assemblies in the reactor are reolaced.
The basic scheme of the fuel reloading was placement of the fresh fuel assemblies on the side of the reactor core, and their relocation towards the centre of the reactor core during fuel assembly replacements in the individual years. In an economical point of view (use of fuel), this scheme was not an ideal one. In addition, the fresh fuel assemblies provide a higher performance in the reactor core, and their location on the side of the reactor core was not ideal even in viewpoint of the radiation load of the reactor vessel (high neutron flows contribute to degradation of the reactor vessel). Improved fuel parameters enabled transition from a three- to four-year fuel cycle in 1997, and since 2003 the five-year cycle has also been successively started up.
Regulation of the reactor performance
In the course of the reactor operation the nuclear fuel is burning up (uranium contents in the fuel is reduced), and this needs to be compensated by reducing the boric acid contents (the so-called absorbent – i.e. the substance that absorbs neutrons) in cooling water. This process is called the long-term changes of the reactor performance.
The short-term fast changes of the reactor performance are carried out by means of a group of seven control rod assemblies. The control rod assembly consists of the fuel part which is the same as the ordinary fuel assembly, and the absorbing part which is of the same shape but made of boron steel.
Each control rod assembly is linked, by means of the inserted rod, with the electric drive located on the cover of the reactor vessel.
When the control rod assemblies are inserted downwards, the fuel part is lifted out of the reactor core, and the absorbing part of the control rod assembly is inserted into its place. This results in the increased absorption of neutrons and the reactor performance is reduced. On the contrary, when the control rod assembly is moved upwards, the reactor performance is increased.
Fast shut-down of the reactor
The fast shut-down of the reactor, or scram, is a fast discontinuation of the fission reaction, and it is one of the essential demands made on nuclear energy. For this purpose, the reactor is equipped with a safety protection system. This system consists of 37 control rod assemblies with the appropriate electronic circuits that put the system automatically into operation in case of an inadmissible exceeding of the permissible parameters and technological condition of the primary or secondary circuit.
This system can also be put into operation upon operator intervention by pressing the push-button at the unit control room.
In case of meeting the conditions for activation of the safety protection system, the power supply for all electric drives that maintain the control rod assemblies in the upper positions is discontinued. After discontinuation of the electric drives power supply, all the control rod assemblies start moving downwards under their own mass into the reactor core, and the fission reaction is terminated within 12 seconds.
Protection against radiation
Protection against radiation in nuclear facilities situated in the territory of the Czech Republic is defined in Act No. 18/1997 Coll. on peaceful utilisation of nuclear energy and ionising radiation (Atomic Act) and Enforcing Regulation No. 184/1997 Coll. on the requirements for ensuring radiological protection. Both the Atomic Act and the Enforcing Regulation on radiological protection stipulate requirements for personal and environment protection system against undesired effects of ionising radiation. There are defined basic duties and conditions for realisation of activities related to use of nuclear energy and ionising radiation. Legislation of the Czech Republic in the field of radiation protection consistently comes out of the internationally respected principles of radiation protection, namely recommendations of the International Commission on Radiological Protection and the subsequent basic international standards on radiological protection. The law in the field of radiological protection in the Czech Republic has been harmonised with the appropriate guidelines of the European Union. Assessment of the influence of the gaseous and liquid discharges into the environment forms a part of the safety documentation of the power plant. It is carried out by measurements inside and outside the power plant, and by means of mathematic simulation. The result is compared with the strictly defined permissible (limiting) values. What is measured in the power plant surroundings? Radiation control laboratory workers systematically carry out monitoring of the environment and collect samples of air, soil, water, vegetation and agricultural products, and make expert analysis. Independent measurements are additionally carried out by state supervision bodies too.
Annual levels of discharge activities, released into the air and watercourses, represent only a negligible fraction of the permissible values in the entire course of the power plant operation. Comparison of this indicator with other power plants puts the Dukovany Nuclear Power Plant into the group of the best power plants in the world.
Professional training of personnel
Professional training of personnel
A special attention is drawn to the recruitment of new members of staff to an employment relationship. Apart from selective interviews and tests, potential employees should comply with tests of psychological and health qualification, and complete demanding professional preparation and training. The training and educational centre in the Dukovany Nuclear Power Plant organises preparation of its own employees and, partially, also employees of its suppliers. Basic theoretical preparation is provided in the Preparatory and Educational Centre in Brno. All employees of the power plant obtain the so-called Professional Preparation Standard, i.e. the list of the specified examinations and tests that have to be successfully passed prior to issue of the so-called Job Performance Authorisation. The validity of these tests is time-limited and the employee should periodically repeat these examinations.
Since 1999, a full-scope display simulator is used for training staff members in the power plant. It was developed by a consortium of the Siemens AG Belgatom and CorysTESS companies under participation of experts from the Dukovany Nuclear Power Plant. The full-scope simulator of the replica-tape unit control room developed by the company ORGREZ SC, a. s., Brno in cooperation with the US company GSE Systems, is also used. Since January 2001 this simulator has been included, after granting a licence by the State Office on Nuclear Safety, in the regular staff training system (training of the unit control room operators) in the Dukovany Nuclear Power Plant. The basic training of operators takes more than two years. Based on the successful passing of the state examination, held in the presence of the state examination commission, the Job Performance Authorisation is issued to the operator. Operators must repeatedly participate in simulator training every year, and every two years they must pass the state examinations.
The power plant management systematically and purposefully develops the corporate culture and in-house communication system. Based on its level the culture of safety, attitudes and approaches of employees are also developed. The corporate culture level is periodically diagnosticated and appropriate measures focused on sustainable improvements are introduced. Evaluation of the working performance and working behaviour of all employees also forms a part of the corporate culture.
Specific requirements on utilisation of nuclear energy are defined in the Atomic Act (Act No. 18/1997 Coll.). This act, in accordance with recommendations of the International Atomic Energy Agency and requirements of the European Union regulations, comprehensively covers problems in the use of nuclear energy and ionising radiation in the Czech Republic, and defines the performance and competence of the state administration and state supervision in this field. The act stipulates conditions for securing nuclear safety, protection against radiation, emergency preparedness, physical protection, and defines the state guaranteed regime for safe deposition of radioactive waste and insurance requirements. This act is linked with a number of enforcing regulations, issued by the State Office on Nuclear Safety, stipulating details and methods of implementation of the duties determined by the state. The control documentation and operational regulations of the power plant come out of the Atomic Act, its enforcing regulations and the related legal standards. They have a key importance for reliable and safe operation of the power plant. In the course of recent years, most of the operational regulations were re-elaborated into a new wording that ensures their high usability and quality.
|EDU Stress Tests - Final report||3 373 kB|