Why does cancer develop? development of a cancer cell. The tumor has smooth, clear edges

(RBMK listen)) is a series of nuclear power reactors developed in the Soviet Union. This reactor is a channel, uranium-graphite (graphite-water moderator), boiling type, on thermal neutrons; designed to generate saturated steam pressure of 70 kg / cm?. The heat carrier is boiling water.
Chief designer of the reactor plant: NIKIET, Academician Dollezhal N. A.
Scientific supervisor of the project: IAE them. I. V. Kurchatova, Academician Aleksandrov A. P.
General Designer (LNPP): GSPI-11 (VNIPIET), Gutov A.I.
Chief designer of the turbine plant: HTGZ, "Turboatom", Yu. F. Kosyak
Metal construction developer: TsNIIPSK, N. I. Melnikov
Leading materials science organization:"Prometheus", Kopyrin G.I.
Designer and manufacturer of CPS electromechanical equipment, CTO: Design Bureau of the Bolshevik Plant, Yu. G. Klaas

On this moment the series of these reactors includes three generations.

Series head reactor- Units 1 and 2 of the Leningrad NPP.

History of creation and operation

Central hall of RBMK-1500

(Ignalina NPP)

The reactor The world's first nuclear power plant was precisely the AM-1 water-cooled uranium-graphite channel reactor (Atom Mirny) installed at the Obninsk NPP (1954). The development of technologies for uranium-graphite reactors was carried out at industrial reactors, including "dual" purpose reactors (which, in addition to "military" isotopes, produced electricity): A (1948), AI (PO Mayak), I-1 (1955 year), EI-2 (1958), ADE series (Siberian Chemical Combine). Since the 1960s, the development of purely power reactors of the future RBMK type has begun in the USSR. Some design solutions were tested on experimental power reactors "Atom Mirny Bolshoi": AMB-1 (1964) and AMB-2 (1967), installed at the Beloyarsk NPP.

The development of the RBMK reactors proper began in the mid-1960s and relied to a large extent on extensive and successful experience in the design and construction of industrial uranium-graphite reactors. The main advantages of the reactor plant were seen by the creators in:

maximum application of the experience of uranium-graphite reactors;

well-established links between factories, well-established production of basic equipment;

the state of industry and the construction industry of the USSR;

promising neutronic characteristics (low fuel enrichment).

In general, the design features of the reactor repeated the experience of previous uranium-graphite reactors. New steel fuel channel, assembly of fuel elements from new structural materials - zirconium alloys, and with new form fuel - metallic uranium was replaced by its dioxide, as well as the parameters of the coolant. The reactor was originally designed as a single-purpose reactor - for the production of electrical and thermal energy.

Work on the project began at the IAE (RRC KI) and NII-8 (NIKIET) in 1964. In 1965, the project was named B-190, and its design was entrusted to the design bureau of the Bolshevik plant. In 1966, by decision of the ministerial NTS, work on the project was entrusted to NII-8 (NIKIET), led by Dollezhal.

On April 15, 1966, the head of the Minsredmash, E.P. Slavsky, signed an assignment for the design of the Leningrad nuclear power plant, 70 km in a straight line west of Leningrad, 4 km from the village of Sosnovy Bor. In early September 1966, the design assignment was completed.

On November 29, 1966, the Council of Ministers of the USSR adopted Decree No. 800-252 on the construction of the first stage of the Leningrad NPP, defined the organizational structure and cooperation of enterprises for the development of the design and construction of the NPP.

The first power unit with an RBMK-1000 type reactor was launched in 1973 at the Leningrad NPP.

During the construction of the first power plants there was an opinion in our country that a nuclear power plant is a reliable source of energy, and possible failures and accidents are unlikely, or even hypothetical, events. In addition, the first units were built within the system of medium mechanical engineering and were supposed to be operated by organizations of this ministry. Safety regulations at the time of development either did not exist or were imperfect. For this reason, the first power reactors of the RBMK-1000 and VVER-440 series did not have a sufficient number of safety systems, which required further serious modernization of such power units. In particular, in the initial design of the first two RBMK-1000 units of the Leningrad NPP, there were no hydrocylinders for the emergency reactor cooling system (ECCS), the number of emergency pumps was insufficient, there were no check valves(OK) on distributing-group collectors (RGK), etc. Later, in the course of modernization, all these shortcomings were eliminated.

Further construction of RBMK blocks was supposed to be carried out for the needs of the USSR Ministry of Energy. Taking into account the less experience of the Ministry of Energy in working with nuclear power plants, significant changes were made to the project that increase the safety of power units. In addition, changes were made to take into account the experience of the first RBMKs. Among other things, ECCS hydrocylinders were used, 5 pumps began to perform the function of emergency ECCS electric pumps, check valves were used in the RGK, and other improvements were made. According to these projects, power units 1, 2 of the Kursk NPP and 1, 2 of the Chernobyl NPP were built. At this stage, the construction of RBMK-1000 power units of the first generation (6 power units) was completed.

Further improvement of NPPs with RBMK began with the development of projects for the second stage of the Leningrad NPP (power units 3, 4). The main reason for finalizing the project was the tightening of security rules. In particular, a system of balloon ECCS, ECCS of long-term cooldown, represented by 4 emergency pumps, was introduced. The accident localization system was represented not by a bubbler tank, as before, but by an accident localization tower capable of accumulating and effectively preventing the release of radioactivity in case of accidents with damage to the reactor pipelines. Other changes have been made. The main feature of power units 3, 4 of the Leningrad NPP was the technical decision on the location of the RGC at an altitude higher than the altitude of the core. This made it possible to have a guaranteed filling of the core with water in the event of an emergency water supply to the RGC. Subsequently, this decision was not applied.

After the construction of power units 3, 4 of the Leningrad NPP, which is under the jurisdiction of the Ministry of Medium Machine Building, the design of RBMK-1000 reactors for the needs of the USSR Ministry of Energy began. As noted above, when developing a nuclear power plant for the Ministry of Energy, additional changes were made to the project to improve the reliability and safety of nuclear power plants, as well as increase its economic potential. In particular, when finalizing the second stages of the RBMK, a drum-separator (BS) of a larger diameter was used (the inner diameter was increased to 2.6 m), a three-channel ECCS system was introduced, the first two channels of which were supplied with water from hydraulic cylinders, the third - from feed pumps. The number of pumps for emergency water supply to the reactor was increased to 9 units and other changes were made, which significantly increased the safety of the power unit (in principle, the level of execution of the ECCS met not only the documents in force at the time of the design of the NPP, but also, in many respects, modern requirements). The capabilities of the accident localization system were significantly increased, which was designed to counter an accident caused by a guillotine rupture of a pipeline of maximum diameter (pressure collector of the main circulation pumps(MCP) Du 900). Instead of bubble tanks of the first stages of the RBMK and containment towers of 3.4 units of the Leningrad NPP, two-story containment pools were used at the RBMK of the second generation of the Ministry of Energy, which significantly increased the capabilities of the accident localization system (ALS). The lack of containment was compensated for by the strategy of using a system of tight-strength boxes (TSBs), in which the pipelines of the multiple loop were located. forced circulation coolant. The design of the FPB, the thickness of the walls were calculated from the condition of maintaining the integrity of the premises in case of a rupture of the equipment located in it (up to the pressure manifold of the MCP DN 900 mm). PPB was not covered by BS and steam-water communications. Also, during the construction of the NPP, the reactor compartments were built in a double block, which means that the reactors of the two power units are essentially in the same building (unlike previous NPPs with RBMK, in which each reactor was in a separate building). So the RBMK-1000 reactors of the second generation were made: power units 3 and 4 of the Kursk NPP, 3 and 4 of the Chernobyl NPP, 1 and 2 of the Smolensk NPP (together, together with the 3 and 4 unit of the Leningrad NPP, 8 power units).

A total of 17 power units with RBMK were put into operation. The payback period for serial blocks of the second generation was 4-5 years.

Contribution of NPPs with RBMK reactors to total output of electricity by all NPPs in Russia is about 50%.

Before the accident at the Chernobyl nuclear power plant in the USSR, there were extensive plans for the construction of such reactors, but after the accident, plans to build RBMK power units at new sites were curtailed. After 1986, two RBMK reactors were put into operation: RBMK-1000 at Smolensk NPP (1990) and RBMK-1500 at Ignalina NPP (1987). Another RBMK-1000 reactor of Unit 5 of the Kursk NPP is under construction (~70-80% completion). After the accident at the Chernobyl nuclear power plant, additional research and modernization. At present, RBMK reactors are not inferior in terms of safety and economic performance to domestic and foreign nuclear power plants of the same period of construction. To date, the acceptable level of safety of RBMK has been confirmed at the national level, as well as by international experts.

The development of the concept of a channel uranium-graphite reactor is carried out in the projects of the MKER - Multi-loop Channel Power Reactor.

Characteristics of RBMK

Design

Scheme of a nuclear power plant
with RBMK type reactor

One of the goals in the development of the RBMK reactor was to improve the fuel cycle. The solution to this problem is associated with the development of structural materials that weakly absorb neutrons and differ little in their mechanical properties from stainless steel. Reducing the absorption of neutrons in structural materials makes it possible to use cheaper nuclear fuel with low uranium enrichment (according to the original project - 1.8%).

RBMK-1000

The basis of the RBMK-1000 core is a graphite cylinder 7 m high and 11.8 m in diameter, made of smaller blocks, which acts as a moderator. graphite permeated big amount vertical holes, through each of which passes a pressure pipe (also called technological channel(TC)). The central part of the pressure tube, located in the core, is made of zirconium alloy (Zr + 2.5% Nb), which has high mechanical and corrosion properties, the upper and lower parts of the pressure tube are made of stainless steel. The zirconium and steel parts of the pressure pipe are connected by welded adapters.

When designing RBMK power units, due to the imperfection of the calculation methods, the channel array spacing was not chosen optimally. As a result, the reactor turned out to be somewhat slowed down, which led to positive values vapor reactivity coefficient in the working area, exceeding the fraction of delayed neutrons. Before the accident at the Chernobyl nuclear power plant, the method used to calculate the steam reactivity coefficient curve (BMP program) showed that despite the positive RCC in the area of ​​operating steam content, as the steam content increases, this value changes sign, so that the dehydration effect turned out to be negative. Accordingly, the composition and performance of security systems was designed with this characteristic in mind. However, as it turned out after the accident at the Chernobyl nuclear power plant, the calculated value of the vapor reactivity coefficient in areas with high vapor content was obtained incorrectly: instead of being negative, it turned out to be positive. To change the vapor reactivity coefficient, a number of measures were taken, including the installation of additional absorbers instead of fuel in some channels. Subsequently, in order to improve the economic performance of power units with RBMK, additional absorbers were removed, and in order to achieve the desired neutron-physical characteristics, fuel of a higher enrichment with an additional absorber (erbium oxide) was used.

Each fuel channel has a cassette made up of two fuel assemblies(TVS) - lower and upper. Each assembly includes 18 fuel rods. The TVEL shell is filled with uranium dioxide pellets. According to the original design, the enrichment in uranium 235 was 1.8%, but as experience in operating the RBMK was gained, it turned out to be expedient to increase the enrichment. The increase in enrichment, combined with the use of a burnable poison in the fuel, made it possible to increase the controllability of the reactor, improve safety and improve its economic performance. Currently, a transition to fuel with an enrichment of 3.0% is being carried out.

The RBMK reactor operates according to a single-loop scheme. The coolant is circulated in a multiple forced circulation loop (MPC). In the core, the water cooling the fuel rods partially evaporates and the resulting steam-water mixture enters the separator drums. Separation of steam takes place in the drum separators, which enters the turbine unit. The remaining water is mixed with feed water and is fed into the reactor core with the help of the main circulation pumps (MCP). The separated saturated steam (temperature ~284 °C) under a pressure of 70-65 kgf/cm2 is supplied to two turbogenerators with an electric power of 500 MW each. The exhaust steam is condensed, after which, after passing through the regenerative heaters and the deaerator, it is supplied by feed pumps (FPU) to the MPC.

RBMK-1000 reactors are installed at Leningrad NPP, Kursk NPP, Chernobyl NPP, Smolensk NPP.

5th power unit of the Kursk NPP
(RBMK-1000 3rd generation)

At Unit 5 of the Kursk NPP, which is currently under construction (readiness at the moment is 70-80%), in addition to other measures to improve the RBMK, the design of the graphite stack of the reactor, which has an octagonal cross-section, has a fundamental novelty. By reducing the volume of graphite, the ratio of the fuel fraction to the moderator fraction changes, which has a significant effect on the vapor reactivity coefficient. As a result, with a guaranteed negative vapor coefficient of reactivity, the RBMK-1000 reactor of Unit 5 of the Kursk NPP operates with a minimum ORM, which further increases its economic efficiency. In the future, it is possible to consider the issue of increasing the fuel enrichment for RBMK Unit 5 of the Kursk NPP, which will further improve its economic performance while maintaining high level security.

This block formally belongs to the 3rd generation of RBMK (the 3rd block of the Smolensk NPP also belongs to it), but, according to the depth of the changes made, it would be more correct to attribute it to the “3+” generation.

RBMK-1500

In RBMK-1500, the power was increased by increasing the specific energy intensity of the core by increasing the power of the FC by 1.5 times while maintaining its design. This is achieved by intensifying the heat removal from the fuel rod by using special heat transfer intensifiers (turbulators) in the TVC in the upper part of both fuel assemblies. All together, this allows you to save the previous dimensions and the overall design of the reactor.

RBMK-1500 FA intensifiers should be distinguished from spacer grids installed on each FA in the amount of 10 pieces, which also contain turbulators.

During operation, it turned out that, due to the high unevenness of energy release, periodically occurring increased (peak) powers in individual channels lead to cracking of the TVEL cladding. For this reason, the power was reduced to 1300 MW.

These reactors are installed at the Ignalina NPP (Lithuania).

RBMK-2000, RBMK-3600
RBMKP-2400, RBMKP-4800
(former projects)

By virtue of common feature design of RBMK reactors, in which the core, like cubes, was recruited from a large number elements of the same type, the idea of ​​a further increase in power suggested itself.

RBMK-2000, RBMK-3600

In project RBMK-2000 the increase in power was planned due to an increase in the diameter of the fuel channel, the number of fuel elements in the cassette and the pitch of the tube grid of the FC. At the same time, the reactor itself remained in the same dimensions.

RBMK-3600 was only a conceptual design, little is known about its design features. It is likely that the issue of increasing the specific power in it was solved, like the RBMK-1500, by intensifying the heat removal, without changing the design of its RBMK-2000 base - and, therefore, without increasing the core.

RBMKP-2400, RBMKP-4800

MKER (modern projects)

The MKER reactor plant projects are an evolutionary development of the generation of RBMK reactors. They take into account new, toughened safety requirements and eliminate the main shortcomings of the previous reactors of this type.

The work of MKER-800 and MKER-1000 is based on natural circulation coolant, intensified by water-to-water injectors. MKER-1500 view large sizes and power works with forced circulation of the coolant developed by the main circulation pumps. Reactors of the MKER series are equipped with a double containment - containment: the first is steel, the second is reinforced concrete without creating a prestressed structure. The diameter of the containment of the MKER-1500 is 56 meters (corresponds to the diameter of the containment of the Bushehr NPP). Due to the good balance of neutrons, MKER reactor plants have a very low consumption of natural uranium (for MKER-1500 it is 16.7 g/MWh(e) - the lowest in the world).

Expected efficiency - 35.2%, service life 50 years, enrichment 2.4%.

Advantages

Reduced water pressure in the primary circuit compared to vessel-type VVERs;

Thanks to the channel design, there is no expensive housing;

No expensive and complex steam generators;

There are no fundamental restrictions on the size of the core (for example, it can be in the form of a parallelepiped, as in RBMKP projects);

Independent circuit of the control and protection system (CPS);

Wide opportunities for regular monitoring of the condition of core components (for example, pipes of technological channels) without the need to shut down the reactor, and also

high maintainability;

Easier (compared to vessel VVER) accidents caused by depressurization of the circulation circuit, as well as transient modes caused by equipment failures;

Possibility to form optimal neutron-physical properties of the reactor core (reactivity coefficients) at the design stage;

Insignificant reactivity coefficients for coolant density (modern RBMK);

Replacement of fuel without shutting down the reactor due to the independence of the channels from each other (in particular, it increases the power factor);

Possibility of production of radionuclides for technical and medical purposes, as well as radiation doping of various materials;

Absence (compared to vessel-type VVERs) of the need to use boron regulation;

More uniform and deeper (compared to vessel-type VVERs) nuclear fuel burn-up;

Ability to operate a reactor with a low ORM - operational reactivity margin (modern projects, for example, the fifth power unit of the Kursk NPP under construction);

More cheap fuel due to the lower enrichment, although the fuel load is much higher (the overall fuel cycle uses the reprocessing of spent fuel from

Channel-by-channel regulation of coolant flow rates through the channels, which makes it possible to control the thermal reliability of the core;

Thermal inertia of the core, which significantly increases the reserves before fuel damage during possible accidents;

Independence of the loops of the reactor cooling circuit (in RBMK - 2 loops), which makes it possible to localize accidents in one loop.

Flaws

A large number of pipelines and various auxiliary subsystems require a large number highly qualified personnel;

The need for channel-by-channel regulation of flow rates, which may lead to accidents associated with the termination of the coolant flow through the channel;

Higher load on operating personnel compared to VVER, associated with a large number of units (for example, shut-off and control valves);

A larger amount of activated structural materials due to the large size of the core and the metal consumption of the RBMK, which remain after decommissioning and require disposal.

Operational practice

IAEA, PRIS Database.
Cumulative capacity factor for all operating power units:
RBMK - 69.71%; VVER - 71.54%.
Data from the beginning of the block introduction to 2008.
Russian Federation. Only active blocks.

Accidents at power units with RBMK

The most serious incidents at nuclear power plants with RBMK reactors:

1975 - rupture of one channel at the first block of Leningrad NPP;

1982 - rupture of one channel at the first unit of the Chernobyl nuclear power plant;

1986 - an accident with a massive rupture of channels at the fourth block of the Chernobyl nuclear power plant;

1991 - fire in the engine room of the second block of the Chernobyl nuclear power plant;

1992 - rupture of one channel at the third unit of Leningrad NPP;

The 1982 accident was associated with the actions of operational personnel, who grossly violated the technological regulations.

In the accident of 1986, in addition to personnel violations, there were dangerous properties RBMK, which significantly affected the scale of the accident. After the accident, a lot of scientific and technical work was carried out. The measures taken have eradicated such dangerous properties.

The accident in 1991 in the engine room of the second block of the Chernobyl nuclear power plant was caused by equipment failures that did not depend on the reactor plant. During the accident, as a result of a fire, the roof of the engine room collapsed. As a result of the fire and the collapse of the roof, the pipelines for feeding the reactor with water were damaged, and the steam relief valve BRU-B was blocked in the open position. Despite the numerous failures of systems and equipment that accompanied the accident, the reactor showed good properties self-protection, which prevented heating and damage to the fuel.

1992 - the rupture of one channel at the third unit of Leningrad NPP was caused by a valve defect.

Status for 2010

As of 2010, 11 power units with RBMK are in operation at three nuclear power plants: Leningradskaya, Kurskaya, Smolenskaya. For political reasons (in accordance with Lithuania's obligations to the European Union), two power units at the Ignalina NPP, three power units at the Chernobyl NPP were shut down (another one ceased to exist as a result of an accident). The construction of the RBMK of the third stage is underway at the fifth power unit of the Kursk NPP.

List of abbreviations, RBMK terminology

A3 - emergency protection; core
AZM - emergency protection (signal) for excess power
AZRT - emergency protection of the reactor plant according to technological parameters (system)
Filling station - emergency protection (signal) by the rate of power slew
AR - automatic regulator
ASKRO - automated system control of the radiation situation
NPP - nuclear power plant
BAZ - high-speed emergency protection
BB - pool bubbler
NIR - side ionization chamber
BOU - block cleaning plant
BRU-D - high-speed reducing device with discharge into the deaerator
BRU-K - high-speed reducing device with discharge into the turbine condenser
BS - separator drum
Main control room - block control panel
VIK - high-altitude ionization chamber
VIUB (SIUB) - lead (senior) unit control engineer
VIUR (SIUR) - lead (senior) reactor control engineer
VIUT (SIUT) - lead (senior) turbine control engineer
GPK - main safety valve
MCP - main circulation pump
DKE (r), (v) - energy release control sensor (radial), (altitude)
DP - additional absorber
DREG - diagnostic registration parameters
ZRK - shut-off and control valve
KGO - cladding tightness control (TVEL-s)
KD - fission chamber
KIUM - installed capacity utilization factor
KMPTS - multiple forced circulation circuit
KN - condensate pump
KCTK - control of the integrity of technological channels (system)
LAZ - local emergency protection
LAR - local automatic regulator
IAEA - International Atomic Energy Agency
MPA - maximum design basis accident
NVK - lower water communications
NK - pressure manifold
NSB - unit shift supervisor
NSS - station shift supervisor
ORM - operational reactivity margin (conditional "rods")
OK - check valve
OPB - " General provisions security"
NBY - "Nuclear Safety Rules"
PVK - steam-water communications
PN - feed pump
PPB - tight-strong boxing
PRIZMA - a program for measuring the power of the device
PEN - feeding electric pump
RBMK - high power channel reactor (boiling water)
RGK - distributing-group collector
RZM - unloading and loading machine
RK CPS - working channel of the control and protection system
RP - reactor space
PP - manual regulation
RU - reactor plant
SAOR - reactor emergency cooling system
SB - security systems
ALS - accident localization system
SP - absorber rod
SPIR - purge and cooldown system
SRK - stop and control valve
STK - process control system
CPS - control and protection system
SFKRE - system physical control power distribution
STsK "Skala" - centralized control system (SKALA - control system of the apparatus of the Leningrad Atomic)
TVS - fuel assembly
TVEL - fuel element
TG - turbogenerator
TC - technological channel
USP - short absorber rod (manual)
NF - nuclear fuel
NFC - nuclear fuel cycle
NPP - nuclear power plant

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The power of the reactor is controlled by changing the value of p, so p is usually a function of time. The number of groups m, as a rule, is 6 - 10, sometimes more (depending on the type of reactor), and therefore the classical solution of the system of these equations is difficult task. The system of equations characterizes the reactor kinetics only under simplifying assumptions; however, for most practical applications, the achieved accuracy is quite sufficient.

The power of the reactors here has also been increased to 36,000 tons per year.

Reactor power can be expressed in watts, kilowatts or megawatts. There is a direct relationship between power, average neutron flux and reactor volume.

The power of a reactor largely depends on its cooling system. This is the peculiarity of nuclear reactors as energy sources. Instantaneous power is determined by the number of atoms undergoing fission in 1 second. In some reactors, it is enough to remove the control rods to cause chain reaction, and to stabilize the reaction (when a certain level is reached), reintroduce the rods into the reactor. In this case, of course, it is assumed that the materials from which the reactor is built can withstand the temperature that may arise there, and this depends only on the efficiency of the cooling system.

The power of the reactor as a steam generator is determined by the number of channels and the power of each of them. With these parameters of the channels, the steam productivity of the reactor-steam generator depends on the number of channels. The more of them, the higher the steam productivity, however, the design of the installation and its operation become more complicated.

The reactor power changes from nominal to idle power in less than 2 s. The AC windings of each phase are located on two vertical rods of a separate core.

The power of the reactors Qp0, intended to regulate the voltage at the intermediate substation, is equal to the sum of the reactive powers of the ends of the section adjacent to the substation.

The power of the reactors of the intermediate compensating points of the CP is equal to the sum of the reactive power sinks of the ends of the sections adjacent to the CP.

Chapter 4. Measurement of nuclear-physical parameters of reactors.

4. MEASURING THE NUCLEAR PHYSICAL PARAMETERS OF REACTORS.

4.1. GENERAL INFORMATION ABOUT MEASURING THE NUCLEAR PHYSICAL PARAMETERS OF THE REACTOR.

The main parameters of the operation of a nuclear reactor are: power, the rate of its change (reactivity and period) and the distribution of energy release in the core. According to the indicators of nuclear-physical parameters, the optimal operating modes of the reactor are established.

Power measurement R reactor can be carried out according to thermal parameters (temperature difference t out and t in and coolant flow Q T :

P=sQ T(texit-t in). (4.1)

However, this method has significant drawbacks: large inertia and impossibility of application at low power levels. In addition, the measurement of the coolant flow rate gives a rather large error. Thus, the measurement of thermal power can be used for calculations of technical and economic indicators, but not for reactor control.

Measurement of the nuclear-physical parameters of the reactor is practically free from the above disadvantages. Measuring the neutron flux density in the reactor core makes it possible to measure the reactor power from "zero" to "nominal", because the reactor power level is proportional to the number of neutrons in the core.

Neutron transducer signal I and thermal power of the reactor R related by the approximate expression:

P=K 1 TO 2 TO 3 I, (4.2)

Where K 1 - coupling coefficient between the neutron flux at the location of the transducer and the transducer signal;

K 2 - coupling coefficient between the average neutron flux in the reactor and the neutron flux at the location of the converter;

K 3 - coefficient of connection between the thermal power and the average neutron flux in the reactor.

A distinctive feature of a nuclear reactor as an object of control and management is that its start-up begins with a very low level power. Therefore, power measurement must be carried out in a wide range from the lowest level to a level exceeding the rated power. Coverage of measurements in such a wide range by one instrument is impossible, therefore, several measuring instruments with different sensitivities are used. On fig. Figure 4.1 shows an approximate distribution of reactor power control ranges.

Fig.4.1. Power control ranges.

Allocate the following modes of operation of the reactor.

Shut down reactor when the reactor is in a subcritical state. The minimum power level of a shutdown reactor can be 10 -11 - 10 -10 of the nominal level. The energy release mainly determines the residual γ-radiation.

Reactor start, when the reactor is brought from subcritical to critical. This state corresponds to an increase in power up to 10 -10 - 10 -8 of the nominal. In this mode, the reactor is controlled manually by the operator. The control rods are removed in small steps. The rate of change in reactivity is determined by the given period of reactor acceleration. In this mode, the controls require reliable control of the power and acceleration period.

Conclusion to power. IN In this mode, the reactor power is increased to the level of 1 - 2% of the nominal, from which the heating of the elements begins due to nuclear fission. The control means provide the required rate of ascent and compensation for reactivity changes associated with the heating of the reactor and the rise in power. Particular attention is paid to the transitional modes of operation of all elements of the system.

Work at rated power. IN In this mode, the reactor must meet the requirements of the power system. Control systems provide control and protection of the reactor, compensate for xenon poisoning of the reactor and burnout.

Reactor shutdown. The stop mode is carried out by the controlled introduction of negative reactivity. The reactor power varies from the nominal level to the minimum level corresponding to the shutdown reactor.

The given operating modes of the reactors cause different requirements for the control and protection systems of the reactor. Measuring channels are divided into separate subsystems: launch channels and reactor control channels at energy power levels.

The launch channels control the neutron flux density and the period of the reactor in a subcritical state, when the reactor is brought to a critical state and when the power is raised to (0.1 - 1) nom. Information about the neutron flux is carried out using pulse and current devices with logarithmic scales covering 6-7 orders of change in the neutron flux.

Measuring transducers for monitoring power, period and reactivity are installed outside the active zone. In channel reactors (Fig. 4.2) transducers 3 installed between the reflector 1 reactor 2 and biological protection 4, in vessel reactors - between the vessel and the protection.

At the moment, the series of these reactors includes three generations. The head reactor of the series is the 1st and 2nd units of the Leningrad NPP.

Encyclopedic YouTube

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✪ RBMK-1000 reactor

✪ Power nuclear reactors

✪ Installation of the multiple forced circulation circuit of the RBMK-1000 reactor

✪ Chernobyl NPP - 30 years after the accident ☢ Chernobyl NPP - 30 years after the accident

✪ REM overload

Subtitles

History of creation and operation

The reactor of the world's first nuclear power plant (AM-1 ("Atom Mirny"), Obninsk NPP, 1954) was precisely a uranium-graphite channel reactor with a water coolant. The development of uranium-graphite reactor technologies was carried out at industrial reactors, including “dual” purpose reactors (dual-purpose reactors), which, in addition to “military” isotopes, produced electricity and used heat to heat nearby cities.

Industrial reactors that were built in the USSR: A (1948), AI (PO Mayak in Ozersk), AD reactors (1958), ADE-1 (1961) and ADE-2 (1964) (Mining and chemical plant in Zheleznogorsk), reactors I-1 (1955), EI-2 (1958), ADE-3, ADE-4 (1964) and ADE-5 (1965) ( Siberian chemical plant in Seversk).

The development of the RBMK reactors proper began in the mid-1960s and relied to a large extent on extensive and successful experience in the design and construction of industrial uranium-graphite reactors. The main advantages of the reactor plant were seen by the creators in:

• maximum application of the experience of uranium-graphite reactors;
• well-established links between factories, well-established production of basic equipment;
• the state of industry and the construction industry of the USSR;
• promising neutronic characteristics (low fuel enrichment).

In general, the design features of the reactor repeated the experience of previous uranium-graphite reactors. The fuel channel became new, assemblies of fuel elements from new structural materials - zirconium alloys, and with a new form of fuel - metallic uranium was replaced by its dioxide, as well as coolant parameters. The reactor was originally designed as a single-purpose reactor - for the production of electrical and thermal energy.

Work on the project began at the IAE (RRC KI) and NII-8 (NIKIET) in 1964. In 1965, the project was named B-190, and its design was entrusted to the design bureau of the Bolshevik plant. In 1966, by decision of the ministerial NTS, work on the project was entrusted to NII-8 (NIKIET), led by Dollezhal.

During the construction of the first nuclear power plants in the USSR, there was an opinion that a nuclear power plant is a reliable source of energy, and possible failures and accidents are unlikely or even hypothetical events. In addition, the first units were built within the system of medium mechanical engineering and were supposed to be operated by organizations of this ministry. Safety regulations at the time of development either did not exist or were imperfect. For this reason, the first power reactors of the RBMK-1000 and VVER-440 series did not have a sufficient number of safety systems, which required further serious modernization of such power units. In particular, in the initial design of the first two RBMK-1000 units of the Leningrad NPP, there were no hydrocylinders for the emergency reactor cooling system (ECCS), the number of emergency pumps was insufficient, there were no check valves (OK) on the distributing-group manifolds (RGK), etc. In the future , in the course of modernization, all these shortcomings were eliminated.

Further construction of RBMK blocks was supposed to be carried out for the needs of the Ministry of Energy and Electrification of the USSR. Taking into account the less experience of the Ministry of Energy with nuclear power plants, significant changes were made to the project that increase the safety of power units. In addition, changes were made to take into account the experience of the first RBMKs. Among other things, ECCS hydrocylinders were used, 5 pumps began to perform the function of emergency ECCS electric pumps, check valves were used in the RGK, and other improvements were made. According to these projects, power units 1, 2 of the Kursk NPP and 1, 2 of the Chernobyl NPP were built. At this stage, the construction of RBMK-1000 power units of the first generation (6 power units) was completed.

Further improvement of NPPs with RBMK began with the development of projects for the second stage of the Leningrad NPP (power units 3, 4). The main reason for finalizing the project was the tightening of security rules. In particular, a system of balloon ECCS, ECCS of long-term cooldown, represented by 4 emergency pumps, was introduced. The accident localization system was represented not by a bubbler tank, as before, but by an accident localization tower capable of accumulating and effectively preventing the release of radioactivity in case of accidents with damage to the reactor pipelines. Other changes have been made. The main feature of the third and fourth power units of the Leningrad NPP was the technical decision on the location of the RGC at an altitude higher than the altitude of the active zone. This made it possible to have a guaranteed filling of the core with water in the event of an emergency water supply to the RGC. Subsequently, this decision was not applied.

After the construction of power units 3, 4 of the Leningrad NPP, which is under the jurisdiction of the Ministry of Medium Machine Building, the design of RBMK-1000 reactors for the needs of the USSR Ministry of Energy began. As noted above, when developing a nuclear power plant for the Ministry of Energy, additional changes were made to the project, designed to increase the reliability and safety of nuclear power plants, as well as increase its economic potential. In particular, when finalizing the second stages of the RBMK, a drum-separator (BS) of a larger diameter was used (the inner diameter was increased to 2.6 m), a three-channel ECCS system was introduced, the first two channels of which were supplied with water from hydraulic cylinders, the third - from feed pumps. The number of pumps for emergency water supply to the reactor was increased to 9 units and other changes were made that significantly increased the safety of the power unit (the level of execution of the ECCS complied with the documents in force at the time of designing the NPP. The capabilities of the accident localization system, which was designed to counteract an accident caused by a guillotine rupture, were significantly increased pipeline of maximum diameter (pressure manifold of the main circulation pumps (MCP) Du 900. Instead of bubble tanks of the first stages of the RBMK and localization towers of Units 3 and 4 of the Leningrad NPP, two-story localization pools were used at the RBMK of the second generation of the Ministry of Energy, which significantly increased the capabilities of the localization system accidents (SLA). MCP manifold DN 900 mm). PPB was not covered by BS and steam-water communications. Also, during the construction of the NPP, the reactor compartments were built in a double block, which means that the reactors of the two power units are essentially in the same building (unlike previous NPPs with RBMK, in which each reactor was in a separate building). So the RBMK-1000 reactors of the second generation were made: power units 3 and 4 of the Kursk NPP, 3 and 4 of the Chernobyl NPP, 1 and 2 of the Smolensk NPP (together, together with the 3 and 4 unit of the Leningrad NPP, 8 power units).

A total of 17 power units with RBMK were put into operation. The payback period for serial blocks of the second generation was 4-5 years.

The contribution of NPPs with RBMK reactors to the total electricity generation by all NPPs in Russia is about 50%.

Characteristics of RBMK

Design

One of the goals in the development of the RBMK reactor was to improve the fuel cycle. The solution to this problem is associated with the development of structural materials that weakly absorb neutrons and differ little in their mechanical properties from stainless steel. Reducing the absorption of neutrons in structural materials makes it possible to use cheaper nuclear fuel with low uranium enrichment (according to the original project - 1.8%). Later, the degree of uranium enrichment was increased.

RBMK-1000

Each fuel channel has a cassette made up of two fuel assemblies(TVS) - lower and upper. Each assembly includes 18 fuel rods. The fuel element cladding is filled with uranium dioxide pellets. According to the original design, the uranium-235 enrichment was 1.8%, but, as experience in operating the RBMK was gained, it turned out to be advisable to increase the enrichment. The increase in enrichment, combined with the use of a burnable poison in the fuel, made it possible to increase the controllability of the reactor, improve safety and improve its economic performance. At present, a transition has been made to fuel with an enrichment of 2.8%.

The RBMK reactor operates according to a single-loop scheme. The coolant is circulated in a multiple forced circulation loop (MPC). In the core, the water cooling the fuel rods partially evaporates and the resulting steam-water mixture enters the separator drums. Separation of steam takes place in the drum-separators, which enters the turbine unit. The remaining water is mixed with feed water and is fed into the reactor core with the help of the main circulation pumps (MCP). The separated saturated steam (temperature ~284 °C) under a pressure of 70-65 kgf/cm 2 is supplied to two turbogenerators with an electric power of 500 MW each. The exhaust steam is condensed, after which, after passing through regenerative heaters and a deaerator, it is fed to the KMPC using feed pumps (FPUs).

RBMK-1000 reactors are installed at the Leningrad NPP, Kursk NPP, Chernobyl NPP, Smolensk NPP.

Chernobyl accident

RBMK-1500

In RBMK-1500, the power was increased by increasing the specific energy intensity of the core by increasing the power of the FC by 1.5 times while maintaining its design. This is achieved by intensifying the heat removal from the fuel elements by using special heat transfer intensifiers (turbulators) in the TVC in the upper part of both fuel assemblies. All together, this allows you to save the previous dimensions and the overall design of the reactor.

During operation, it turned out that due to the high unevenness of energy release, periodically occurring increased (peak) powers in individual channels lead to cracking of the fuel cladding. For this reason, the power was reduced to 1300 MW.

These reactors were installed at the Ignalina NPP (), and were planned for installation according to the original design of the Kostroma NPP.

RBMK-2000, RBMK-3600, RBMKP-2400, RBMKP-4800, (former projects)

Due to the general design features of the RBMK reactors, in which the core, like cubes, was recruited from a large number of the same type of elements, the idea of ​​a further increase in power suggested itself.

RBMK-2000, RBMK-3600

In project RBMK-2000 the increase in power was planned due to an increase in the diameter of the fuel channel, the number of fuel rods in the cassette and the pitch of the FC tube sheet. At the same time, the reactor itself remained in the same dimensions.

RBMK-3600 was only a conceptual design, little is known about its design features. Probably, the issue of increasing the specific power in it was solved, like the RBMK-1500, by intensifying the heat removal, without changing the design of its RBMK-2000 base - and, therefore, without increasing the core.

RBMKP-2400, RBMKP-4800

They differ from all RBMKs in their active zone in the form of a rectangular parallelepiped and the presence of stainless steel overheating channels. The steam temperature in RBMKP-2400 and RBMKP-4800 is 450 degrees Celsius [ ] .

MKER (modern projects)

Expected efficiency - 35.2%, service life 50 years, enrichment 2.4%.

Advantages

Operational practice

Crash 1982, according to internal analysis chief designer (NIKIET), was associated with the actions of operational personnel who grossly violated the technological regulations.