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Chernobyl: Part 1. Description of the Chernobyl NPP with RBMK-1000 reactors.

Information on the Chernobyl accident and its consequences, prepared for the IAEA Report №1 (INSAG-1)


S U D E F G H I N
Flash drive
0. Introduction
1. Description of the Chernobyl NPP with RBMK-1000 reactors.
2. Chronology of the accident.
3. Analysis of the development of the accident on a mathematical model.
4. Causes of the accident.
5. Preventing the development of the accident and reducing its consequences.
6. Control of radioactive pollution of the environment and public health.
7. Recommendations for improving the safety of nuclear power.





1. DESCRIPTION OF THE CHERNOBYL NPP WITH RBMK-1000 REACTORS

1.1. Project data
1.2. Description of the reactor unit of the fourth unit of ChNPP
1.3. Basic physical characteristics of the reactor
1.4. Security systems
1.5. Description of the site of the Chernobyl NPP and its location


1.1. Project data
The design capacity of the Chernobyl NPP is 6 GW, as of January 1, 1986, the capacity of the four NPP units is 4 GW.

1.2. Description of the reactor unit of the fourth unit of ChNPP
The main design features of RBMK reactors are:
- vertical channels with fuel and coolant that allow local fuel reloading while the reactor is operating;
- fuel in the form of bundles of cylindrical fuel elements of uranium dioxide in zirconium shells;
- graphite retarder between channels;
- light water boiling coolant in the circuit of multiple forced circulation (MFCC) with direct steam supply to the turbine.

The RBMK-1000 with a thermal capacity of 3,200 MW (Fig. 1) is equipped with two identical cooling loops; 840 parallel vertical channels with fuel assemblies are connected to each loop. The cooling loop has four parallel main circulation pumps (MCPs): three operating, supplying 7000 t / h of water each with a pressure of ~ 1.5 MPa, and one standby.

The control and protection system (CPS) of the reactor is based on the displacement of 211 solid absorber rods in dedicated channels cooled by an autonomous circuit. The system provides: automatic maintenance of a given power level; a rapid decrease in power by the rods of automatic regulators (AP) and manual regulators (PP) according to the signals of failure of the main equipment; emergency termination of a chain reaction with emergency rods (A3) due to impulses of dangerous deviations of unit parameters or equipment failures; compensation of changes in reactivity during heating and power output; regulation of energy release in the core.
RBMKs are equipped with a large number of independent regulators that, when activated, AZ are introduced into the core at a speed of 0.4 m / s. The low speed of the regulators is compensated by their number.

The CPS includes subsystems of local automatic control (LAR) and local emergency protection (LAZ). Both work on signals from in-core ionization chambers. LAR automatically stabilizes the main harmonics of the radial-azimuthal distribution of energy release, and LAZ provides the A3 reactor from exceeding the specified power of fuel assemblies in its individual zones. To regulate the high-altitude fields, shortened absorber rods introduced into the zone from below (24 pcs.) Are provided.

In addition to the CPS, the RBMK-1000 provides for the following main monitoring and control systems:
- physical control of the energy release field along the radius (over 100 channels) and along the height (12 channels) using direct charge sensors;
- starting control (reactimeters, starting removable chambers);
- control of water flow through each channel by ball flow meters;
- control of the tightness of the fuel claddings by the short-lived activity of volatile fission products in steam-water communications at the outlet of each channel; activity is detected sequentially in each channel in the corresponding optimal energy ranges (“windows”) of photomultipliers moved by a special trolley from one communication to another;
- monitoring the integrity of the pipe channels by humidity and temperature of the gas washing the channels.

All data goes to the computer. Information is given to operators in the form of deviation signals, readings (on call) and registrar data.
The PBMK-1000 power units operate mainly in the base mode (at constant power). Due to the high power of the unit, the full automatic shutdown of the reactor occurs only when the output levels of power, pressure or water in the separator exceed the permissible limits, general de-energizing, two turbogenerators or two MCPs are disconnected, the feedwater flow drops by more than 2 times, the gap is broken to a full cross section pressure collector MCP with a diameter of 900 mm.


Fig. 1. Cross-section of the main building of nuclear power plants with RBMK-1000, including the localization zone.
List of main equipment of the main building of the NPP


1.3. Basic physical characteristics of the reactor
The RBMK-1000 nuclear power reactor is a heterogeneous channel thermal reactor, which uses low-enriched 235U uranium dioxide as a fuel, graphite as a moderator and boiling light water as a coolant.
Below are the main characteristics of the reactor:

Thermal power, MW .............................................. ......................................... 3200
Fuel enrichment,% .............................................. ............................................. 2.0
Uranium mass in fuel assembly, kg ............................................ ................................................. 114.7
Number / diameter of fuel rods in fuel assembly, mm .......................................... ............................... 18 / 13.6
Burnout depth, MW-days / kg ......................................... ..................... 20
Coefficient of uneven energy release:
radius ……………………………………………………………………………… ..1.48
in height ………………………………………………
Maximum rated power of the channel, kW ............................................ ............. 3250
Steam reactivity coefficient p at the operating point,% -1 by steam volume ... 2.0-10-4
Rapid power coefficient of reactivity aw at the operating point, MW-1 ..- 0.5 10-6
Temperature coefficient of fuel аt, С -1 .......................................... ..........- 1,2 10 -4
Temperature coefficient of graphite ac, C -1 .......................................... ........... 6 10 -5
Minimum efficiency of control rods,% …………………………… ....... 10.5
The effectiveness of the rods of the RR,% .........................7.5
The effect of replacing (on average) burned-up fuel assemblies with fresh one,% …………………… 0.02


An important physical characteristic from the point of view of control and safety of the reactor is the quantity called the operational reactivity margin, that is, a certain number of control rods immersed in the active zone that are in the area of ​​high differential efficiency. It is determined by recalculation for fully immersed control rods. Reactivity margin for RBMK-1000 is assumed to be 30 PP bars. At the same time, the input speed of negative reactivity when triggered by A3 is w / s (c is the fraction of delayed neutrons), which is sufficient to compensate for the positive effects of reactivity.

The dependence of the effective multiplication factor on the density of the coolant in the RBMK is largely determined by the presence of various absorbers in the core. At the initial loading of the core, which includes ~ 240 boron-containing additional absorbers, dehydration leads to a negative reactivity effect. At the same time, a slight increase in steam content at rated power with a reactivity margin of 30 rods leads to an increase in reactivity (p = 2-10 -40% -1 by volume of steam).

For a boiling water graphite reactor, the main parameters determining its performance and safety in thermal terms are: the temperature of fuel elements, the stock before the heat transfer crisis, and the temperature of graphite.

For RBMK, a set of programs has been developed that allows station computers to carry out operational calculations to ensure the thermal technical reliability of the unit in the mode of continuous fuel overloads at any positions of the shut-off and control valves at the inlet to each channel. This makes it possible to determine the thermal parameters of a reactor at different frequencies of regulation per channel costs, different laws of regulation (on the output steam content or on the reserve to critical power), as well as at different degrees of preliminary throttling of the core.

To determine the energy release fields over the reactor core, the readings of the physical control system based on the in-core neutron flux measurements along the radius and height of the core are used. Along with the testimony of the physical control system, data describing the composition of the core, the power generation of each TC, the position of the control rods, the distribution, the flow of water along the channels of the core, and the readings of the pressure sensors and coolant temperature are also entered into the station computer.
The operating experience of operating RBMKs shows that, with the means of control and regulation available in these reactors, maintaining the temperature regime of the fuel, graphite and stock before the heat transfer crisis at an acceptable level does not cause difficulties.


1.4. Security systems (Fig. 2)
1.4.1. Protective security systems.
The reactor emergency cooling system (ECCS) is a protective safety system and is designed to ensure the removal of residual heat by supplying the required amount of water to the reactor channels in case of accidents accompanied by disturbances in the core cooling. Such accidents include: breaks in large diameter pipe mains, steam pipelines and feedwater pipelines.
The system of protection against over-pressure in the main coolant circuit is designed to ensure the permissible pressure in the circuit by draining steam into the bubbler pool for condensation.
The protection system of the reactor space is designed to maintain the pressure in it at a level no higher than permissible in an emergency situation with a rupture of one TC due to the removal of the vapor-gas mixture from the reactor space to the vapor-gas discharge barrier of the bubbler pool and further into the bubbler pool while extinguishing the chain reaction with means A3 . The ECCS and the cooling system of the reactor space can be used to introduce appropriate neutron absorbers (boron salts and 3He).

1.4.2. Localizing security systems.
The accident localization system (ALS) implemented on the fourth block of the ChNPP is designed to localize radioactive emissions during accidents with decompaction of any pipelines of the reactor cooling circuit, except for steam and water communications, the upper paths of the TC and that part of the standpipe pipes located in the BS room and pipelines of vapor-gas discharges from the reactor space.
The main component of the ALS is a system of sealed rooms, including the following rooms of the reactor compartment:
- durable boxes, located symmetrically with respect to the axis of the reactor and designed for an overpressure of 0.45 MPa;
- rooms of distributing group collectors and lower water communications (these rooms do not allow excess pressure to rise above 0.08 MPa according to the strength conditions of the structural elements of the reactor and are designed for this value).
The premises of the solid boxes and the steam distribution corridor are connected to the water volume of the bubbling condensation device by the steam discharge channels.
The system of shut-off and sealing fittings is designed to ensure the tightness of the accident localization zone by cutting off communications connecting the sealed and unpressurized rooms.
The bubbling and condensation device is designed to condense steam generated during an accident with a decomposition of the reactor circuit, when the main safety valves are triggered and when they leak through them in normal operation.

1.4.3. Providing security systems. Power supply to nuclear power plants.
Electric power consumers at nuclear power plants, depending on the requirements for reliability of power supply, are divided into three groups:
- consumers that do not allow interruption of power from fractions of a second to several seconds in any modes, including the mode of complete disappearance of AC voltage from operating and standby auxiliary transformers, and requiring mandatory power after the operation of the A3 reactor;
- consumers allowing in the same modes a power interruption of tens of seconds to tens of minutes and requiring mandatory power after the operation of the A3 reactor;
- consumers that do not require power in the modes of power failure from working and standby transformers of their own needs, and in normal operation of the unit they allow interruption of power during the transfer from working to standby transformer of their own needs.

1.4.4. Security control systems.
Control security systems are designed to automatically turn on protective devices, localizing and providing security systems and monitoring their operation.

1.4.5. Radiation monitoring system.
The NPP radiation monitoring system is an integral part (subsystem) of the automated control system of NPPs and is designed to collect, process and present information about the radiation situation in the NPP premises and in the external environment, on the state of process media and circuits, on radiation doses of personnel in accordance with the applicable standards. and legislation.

1.4.6. NPP control points.
NPP management is carried out at two levels: station and block. All devices ensuring the safety of nuclear power plants are controlled at the block level.


1.5. Description of the site of the Chernobyl NPP and its location
1.5.1. The Chernobyl nuclear power plant is located in the eastern part of a large region, called the Belarusian-Ukrainian Polesie, on the banks of the r. Pripyat, flowing into the Dnieper.
At the beginning of 1986, the total population in the 30-kilometer zone around the NPP was ~ 100 thousand people, of which 49 thousand lived in the city of Pripyat, located to the west of the three-kilometer sanitary-protective zone of the nuclear power plant, and 12.5 thousand. - in the district center of Chernobyl, located 15 km south-east of the nuclear power plant.


Fig. 2. The section for the reactor compartment of nuclear power plants with RBMK-1000, included a localization zone (for position designations, see Fig. 1)


1.5.2. Description of the site of the nuclear power plant and its construction.
The first stage of the ChNPP (two power units with RBMK-1000) was built in 1970-1977, and by the end of 1983. construction of two power units of the second stage was completed at the same site. 1.5 km southeast of this site in 1981. The construction of two more power units with the same reactors was started (the third stage of the NPP).
Southeast of the NPP site directly in the valley. Pripyat has built a 22 km2 bulk cooling pond that provides cooling for turbine condensers and other heat exchangers of the first four power units. The normal water level in the cooling pond was set at 3.5 m below the mark of the NPP site layout.

1.5.3. Data on the number of personnel at the NPP site at the time of the accident.
On the night of April 25-26, 1986, 176 people were on the site of the first and second stages of the Chernobyl NPP - duty personnel, as well as employees of various workshops and repair services.
In addition, 268 builders and installers worked on the site of the third stage of the NPP in the night shift.

1.5.4. Information about the equipment at the site, operating in conjunction with the damaged reactor, and the equipment used in the process of emergency response.
Each stage of the Chernobyl NPP consists of two power units having common systems of special water treatment and auxiliary facilities at the industrial site, which include: storage of liquid and solid radioactive waste; open switchgear; gas industry; backup diesel generator sets; hydrotechnical and other facilities.

The liquid radioactive waste storage facility, built as part of the second stage of the nuclear power plant, is intended for the reception and temporary storage of liquid radioactive waste generated during the operation of the third and fourth units, as well as the reception of operational wash water and return them for processing. Liquid radioactive waste comes from the main building through pipelines laid on the lower tier of the overpass, and solid radioactive waste is fed into the storage facility along the upper corridor of the overpass by electric vehicles.

The backup diesel power station (RDES) is an autonomous emergency source of power supply for systems important to the safety of each unit. Three diesel generators with a unit capacity of 5.5 MW are installed on each RDES of the third and fourth units. To ensure the operation of NDES, intermediate and basic diesel fuel stores, pumping fuel pumps, emergency fuel and oil discharge tanks are provided.

To provide technical water to responsible consumers requiring uninterrupted water supply, separate pumping stations of the third and fourth units with backup power supply from diesel generators are provided.

On April 25, 1986, all four power units of the first and second stages and auxiliary systems and facilities of the industrial site associated with their normal operation were operating.