NUCLEAR DEVELOPMENT AND PROLIFERATION

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP09-00997R000100260001-0
• Release Decision: RIFPUB
• Original Classification: U
• Document Page Count: 370
• Document Creation Date: January 4, 2017
• Document Release Date: July 15, 2013
• Sequence Number: 1
• Publication Date: October 7, 1986
• Content Type: REPORT

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JPRS-TND-86-009-L
7 OCTOBER 1986
FBIS
FOR OFFICIAL USE ONLY
Worldwide Report
NUCLEAR DEVELOPMENT
AND
PROLIFERATION
USSR STATE COMMITTEE REPORT
ON CHERNOBYL NUCLEAR POWER PLANT ACCIDENT
FOREIGN BROADCAST INFORMATION SERVICE
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JPRS-TND-86-009-L
7 OCTOBER 1986
WORLDWIDE REPORT
NUCLEAR DEVELOPMENT AND PROLIFERATION
USSR STATE COMMITTEE REPORT
ON CHERNOBYL NUCLEAR POWER PLANT ACCIDENT
Vienna THE ACCIDENT AT THE CHERNOBYL NUCLEAR POWER PLANT AND ITS
CONSEQUENCES in English 25-29 Aug 86
['Working Document' compiled by USSR State Committee on the Utili-
zation of Atomic Energy for the IAEA Experts'. Post Accident Review
Meeting of 25-29 Aug 19.86 in Vienna, consisting of Part I, General
Material and Part II, Annexes.]
[This document and annexes received with no translation attribution.
For related material see JPRS-TND-86-004-L of 5 June 1986 and
JPRS-UPS-86-038 of 6 August 1986]
CONTENTS
PART I. GENERAL MATERIAL
Preface 1
Introduction 2
1. Description of the Chernobyl' Nuclear Power Station With
RBMK-1000 Reactors 4
2. Chronological Account of How the Accident Evolved 15
3. Analysis of the Accident Using a Mathematical Model ' 18
4. Causes of the Accident 22
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5. Priority Measures for Improving the Safety of Nuclear Power
Plants With RBMK Reactors 24
6. Containment of the Accident and Alleviation of Its Consequences 25
7. Monitoring of Environmental Radioactive Contamination and
Health of the Population 34
8. Recommendations for Improving Nuclear Power Safety 42
9. The Development of Nuclear Power in the USSR 45
PART II. ANNEXES
Annex 1. Water-Graphite Channel Reactors and Operating Experience
With RBMK Reactors 57
Annex 2. Design of the Reactor Plant 62
Annex 3. Elimination of the Consequences of the Accident and
Decontamination 248
Annex 4. Estimate Of the Amount, Composition and Dynamics of
the Discharge of Radioactive Substances From the
Damaged Reactor 253
Annex 5. Atmospheric Transport and Radioactive Contamination of
the Atmosphere and of the Ground 273
Annex 6. Expert Evaluation and Prediction of the Radioecological
State of the Environment in the Area of the Radiation
Plume From the Chernobyl' Nuclear Power Station
(Aquatic Ecosystems) 289
Annex 7: Medical-Biological Problems 297
/6091
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PART I. GENERAL MATERIAL
PREFACE
This material is taken from the conclusions of the Government
Commission on the causes of the accident at the fourth unit of the Chernobyl'
nuclear power plant and was prepared by a team of experts appointed by the
USSR State Committee on the Utilization of Atomic Energy. The members of this
team were:
Abagyan A.A.
Asmolov V.G.
Gus'kova A.K.
Demin V.F.
ll'in L.A.
lzraehl' Yu.A.
Kalugin A.K.
Konviz V.S.
Kuz'min 1.1.
Kunzevich A.D.
Legasov V.A.
Malkin S.D.
Mysenkov A.I.
Pavlovskij O.A.
Petrov V.N.
Pikalov V.K.
Protsenko A.N.
Ryazantsev E.P.
Sivintsev Yu.V.
Sukhoruchkin V.K.
Tokarenko V.F.
Khrulev A.A.
Shakh 0.Ya.
During preparation of this document, material was used from the
following organizations: the I.V. Kurchatov Institute of Atomic Energy, the
Scientific ' Research and Design Institute for Power Technology, the
V.G. Khlopin Radium Institute, the S.Ya. Zhuk 'Hydroproject' Institute, the
All-Union Scientific Research Institute for Nuclear Power. Plants, the
Institute of Biophysics, the Institute of Applied Geophysics, the USSR State
Committee on the Utilization of Atomic- Energy, the USSR State .Committee on
Hydrometerology and Environmental Protection', the Ministry of Health, the USSR
State Nuclear Power Supervisory Board, the Ministry of Defence, the Main Fire
Protection Directorate of the Ministry of Internal Affairs and the USSR
Academy of Sciences.
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INTRODUCTION
On 26 April 1986 at 1.23 a.m. an accident occurred at the fourth unit
of the Chernobyl' Nuclear Power Plant which resulted in the destruction of the
reactor core and part of the building in which it was housed.
,The accident took place prior to shut-down of the unit for planned
maintenance dUring, operating mode tests on one of the turbogenerators. There
was a sudden power surge on the reactor leading to the destruction of the
reactor and the release into the atmosphere of part of the radioactive 1
products which had accumulated in the core. ,
t
During the accident the nuclear reaction in the fourth unit was /
stopped. The fire which occurred was extinguished and work was begun to limit
and eliminate the consequences of the accident.
The population from the areas in the immediate vicinity of the nuclear
power plant and from a 30 km-radius zone around the plant was evacuated.
In view of the extraordinary nature of the Chernobyl' accident, an
operational team headed by the President of the USSR Council of Ministers,
N.I. Ryzhkov, was organized in the Politburo of the CPSU Central Committee to
co-ordinate the activities carried out by the ministries and other state
departments to eliminate the consequences of the accident and to assist the
population. A government commission was set up to study the causes of the
accident and to implement the requisite emergency and rehabilitation
measures. The necessary scientific, technical and economic resources of the
Soviet Union were mobilized.
Representatives of the IAEA were invited to the USSR and given the
opportunity to familiarize themselves with the state of affairs at the
Chernobyl' power plant and the measures taken to control the accident. They
informed the world community of their evaluation.
Governments of a number of countries and many governmental, public and
private organizations and individuals from different countries made offers of
assistance to various Soviet organizations to help eliminate the consequences
of the accident. Some of these offers were accepted.
During the thirty years of its development, nuclear power has occupied
an I important place in world energy-production and on the whole has
demonstrated a very good record of safety for mankind and the environment. It
is impossible to envisage the future of the world economy without nuclear
power. However, its further development must be accompanied by still greater
scientific and technical efforts to guarantee operational reliability and
safety. -
The Chernobyl' accident resulted resulted from a combination of several unlikely
events. The Soviet Union is drawing the appropriate conclusions from the
accident.
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Abandonment of nuclear energy sources would require a significant
increase in the extraction and consumption of organic fuels. This would
undoubtedly increase the risk of disease for mankind, and increase the
destruction of waters and forests as a' result of the constant release of
harmful chemical substances into the biosphere.
In addition to its advantages as a source? of energy and as a means of
conserving natural resources, the world-wide development of nuclear power also
has inherent dangers of an international character. These include
transboundary transfers of radioactivity, particularly in the event of serious
radiation accidents, and the proliferaton of nuclear weapons,- the danger of
international terrorism and the specific danger represented by nuclear
facilities in the event of war. All this emphasizes the crucial need for
close international co-operation in the development of nuclear power and in
ensuring its safety.
That is the reality of the situation.
The fact that the contemporary world is full of potentially dangerous
industrial production processes significantly aggravating the consequences of
military actions gives a new perspective to the senselessness and
inadmissibility of war in today's world.
In his speech on Soviet television on 14 May, M.S. Gorbachev said:
"For us the indisputable lesson of Chernobyl' is that, with the further
development of the scientific and technical revolution, questions of the
reliability and safety of technology, questions of discipline, order and
organization acquire paramount importance. The strictest possible
requirements will have to be applied everywhere and to everything.
"Furthermore we consider that co-operation within the International
Atomic Energy Agency should be further enhanced."
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1. DESCRIPTION OF THE CHERNOBYL' NUCLEAR POWER STATION WITH RBMK-1000
REACTORS
1.1. Design data
The design power output of the Chernobyl' nuclear power plant is 6 GW;
as of 1 January 1986 the power of the four operating units of the station Was
4 GW. The third F and fourth units belong to the second construction stage of
the Chernobyl' nuclear power plant and to the second generation of plants of
that type.
1.2. Description of the reactor in the fourth unit of the Chernobyl' nuclear
power plant
The chief design features of RBMK reactors are the following:
(1) Vertical channels containing the fuel and coolant, enabling local
refuelling while the reactor is in operation;
(2) Fuel in the form of bundles of cylindrical fuel elements made of
uranium dioxide in ziftonium tube-type cladding;
(3) A graphite moderator between the channels;
(4) A boiling light-water coolant in the multiple forced circulation
circuit (MFCC), with direct steam feed to the turbine.
These design features, as a group, determine all the main
characteristics of the reactor and the nuclear power 'plant both as regards its
merits, which include: the absence of cumbersome pressure vessels which are
difficult to- manufacture and limit the reactor's- unit power and production
base; absence of a complex and costly 'steam generator; the possibility of
continuous refuelling and a good neutron balance; a flexible fuel cycle easily
adapted to the fluctuations of the fuel market; the possibility of nuclear
steam superheating; high thermal reliability and durability of the reactor
through channel-by-channel control of coolant flow; channel failure detection,
monitoring of the parameters and coolant activity in each channel and on-load
replacement of leaking assemblies; and as regards its shortcomings: .the
possibility that there may be a positive void coefficient of reactivity due to
the presence of a phase transition in the coolant, which governs the behaviour
of the neutron-flux-determined power during accidents; high sensitivity of the
neutron field to reactivity perturbations of different kinds, requiring a
complicated control system to stabilize the power' density distribution in the
core; complex branching of the coolant delivery and removal system for each
channel; a large Amount of heat energy accumulating in the metal structure's,
fuel elements and graphite structure; 'and slightly radioactive steam in the
turbine.
The RBMK-1000 reactor with a power output of 3200 MW(th) (Fig. 1) is
equipped with two identical cooling loops; to each loop are joined
840 parallel vertical _channels containing the fuel assemblies.
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The cooling loop has four parallel main circulation pumps (three of -
which are operational and feed 7000 t/h of water at a pressure of ?:1.5 MPa,
while one is redundant).
The water in the channels is heated to boiling point and partially
evaporates. The steam-water mixture with a mean mass steam quality of 14% is
led off through the top of the channel and steam-water communication line to
two horizontal gravity-type separators. The dry steam (less than 0.1%
moisture content) separated in them is fed at a pressure of 7 MPa from each
separator via two steam pipes to two turbines with an output of 500 141.1(e)
each, (all eight steam pipes of the four separators are joined in a common
"ring"), while the water, after mixing with the steam condensate, is fed
through 12 downcomers to the section header of the main circulation pumps.
The condensate from the spent steam in the turbines' is recycled through
the separators by feed pumps to the top of the downcomers, thereby subcooling
the water to saturation temperature at the main circulation pump inlet.
As a whole, the reactor consists of a set of vertical fuel and coolant
channels inserted into cylindrical openings in the graphite columns, and in
the top and bottom shielding plates. A light cylindrical cowling encloses the
space occupied by the graphite structure.
This structure consists of graphite blocks assembled in the form of
columns, with a square cross-section and cylindrical axial openings. It rests
on a bottom plate, which transmits the weight of the reactor to a concrete
vault.
About 5% of the reactor power is released in the graphite through the
slowing-down of neutrons and absorption of gamma-quanta. To reduce thermal
resistance and prevent oxidation of the graphite, the cavity in the stack is
filled with a slowly circulating mixture of helium and nitrogen, which serves
at the same time to monitor the integrity of the channels on the basis of
variations in moisture content and temperature of tha gas.
Below the bottom plate and above the top plate there are spaces for
laying the coolant pipes along the routes from the drum separators and
distributing headers to each channel.
The robot, i.e. the refuelling machine, after removal of the relevant
section of floor and lining up with the channel co-ordinates, couples onto the
head of the channel, equalizes its own pressure and the channel pressure,
unseals the channel, removes the burnt-up fuel assembly and replaces it with a
fresh one, reseals the channel, uncouples and transports the spent assembly to
the cooling pond. For as long as the refuelling machine is joined to ,the fuel
channel cavity, a small flow of clean water is fed from it, through the
thermohydraulic sealing, into the fuel channel, thereby creating a "barrier"
to the penetration of hot radioactive water into the refuelling machine from
the channel.
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The control and protection system (CPS) of the reactor is based on the
movement of 211 solid absorber rods in specially separated channels cooled by
water from an autonomous circuit. The system ensures: automatic maintenance
of a set power level; rapid reduction in power by the automatic control rods
and radial controllers on the basis of signals indicating main equipment
failure; emergency stoppage of the chain reaction by the scram rods on the
basis of signals indicating dangerous deviations of the unit parameters or
equipment failure; compensation for reactivity fluctuations when the reactor
is heated up and brought up to power; and control of the power density
distribution through the core.
RBMK reactors are fitted with a large number of independent regulators,
which are inserted into the core at a rate of 0.4 m/s when the emergency
protection system is triggered. The comparatively slow motion of the
regulators is offset by their large number.
The CPS includes sub-systems for local automatic control and local
emergency protection working on the basis of signals from in-core ionization
chambers. The local automatic control system automatically stabilizes the
principal harmonics of the radial-azimuthal power density distribution, while
the local emergency protection system ensures that the reactor, is protected in
an emergency against the subassemblies exceeding the set power in different
regions of the core. To regulate the vertical fields there are shortened
absorber rods, inserted into the core from below (24 rods).
Apart from the CPS, the RBMK-1000 reactor has the following main
monitoring and control systems:
(1) System for physical monitoring of the radial power density field
(more than 100 channels) and the vertical power density field
(12 channels), using direct-charge sensors;
(2) System for monitoring startup (reactivity meters, removable startup
ionization chambers);
(3) System for monitoring water flow through each channel by means of
ball-type flowmeters;
(4) System for fuel failure detection from the short-lived activity of
volatile fission products in the steam-water communication lines at
the outlet from each channel; the activity is detected successively
in each channel over the corresponding optimal energy ranges
("windows") by means of a photomultiplier moved by a special
carriage from one steam-water pipe to another;
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(5) System for montoring channel tube integrity from the moisture
content and temperature of the gas flushing the channels.
All the data are fed to computers. The information is issued to the
__operators in the form of deviation signals, readings (when called for) and
recorder data.
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RBMK-1000 units are used predominantly for base-load operation (at
constant power).
In view of the high power of the unit, the reactor is shut down
automatically only when the readings. for power, pressure and water level, in
the separator go outside permissible limits;, when there is a total loss of
current.;when two turbogenerators or two main circulation pumps cut out at
once;whenthere is a drop of more than a.factor of two in the feedwater flow;
or when a rupture occurs over the whole cross-section of the 900 mm diam. main
circulation pump pressure header. In other cases where the equipment fails,
provision is Made only for an automatically controlled drop in power (to a
level corresponding to the power of the equipment still operating).
Principal physical characteristics of the reactor
The RBMK-1000 nuclear power reactor is a heterogeneous channel-type
thermal reactor in which uranium dioxide slightly enriched in 235U is used
as fuel, graphite is used as moderator and boiling light water is used as
coolant. The reactor has the following principal characteristics:
Thermal power 3200 MW
Fuel enrichment 2.0%
Mass of uranium in fuel assembly 114.7 kg
Number/diameter of fuel elements in
a fuel subassembly 18/13.6 mm
Fuel burnup 20 MW.d/kg
Coefficient of nonuniformity in radial
power density 1.48
Coefficient of nonuniformity in vertical
power density 1.4
Maximum design channel power 3250 kW
Isotopic composition of unloaded fuel:
-Uranium-235 4.5 kg/t
Uranium-236 2.4 kg/t
Plutonium-239 2.6 kg/t
Plutonium-240 1.8 kg/t
Plutonium-241 0.5 kg/t
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Void coefficient of reactivity aq, 2.0 x 10-4 vol.%
at working point steam
Fast power coefficient of reactivity aw
at working point -0.5 x 10-4/MW
Temperature coefficient of fuel aT -1.2 x 10-5/?C
Temperature coefficient Of graphite ac 6 x 10-5/0C
Minimum "weight" of CPS rods, AK 10.5%
Worth of manual control rods, AK 7.5%
Effect of replacing spent fuel
by fresh fuel (average) 0.02%
An important physical characteristic from the standpoint of reactor
control and safety is a quantity known as the operative reactivity margin or
excess reactivity. This is defined in terms of a certain number of CPS rods
inserted into the core in the region of high differential worth, for fully
inserted rods.
The excess reactivity for RBMK-1000 reactors is taken as equalivalent
to 30 manual regulating rods. The rate of insertion of negative reactivity,
when the emergency protection system is triggered is 1 0/s (0 is the
fraction of delayed neutrons), which is sufficient to compensate for positive
reactivity effects.
The nature of the relationship between the effective multiplication
factor and the coolant density in RBMK reactors is largely determined by the
different types of absorbers in the core. With the initial loading of the
emergency protection system, which comprises about 240 additional absorbers
with boron, loss of water draining leads to a negative reactivity effect.
At the same time, a slight increase in steam quality at rated power,
with a reactivity margin of 30 rods, results in a reactivity increase
(p = 2 x 10-4/vol.% steam).
ln the case of a boiling water graphite-moderated reactor, the min
parameters relevant to operational reliability and thermal safety are the
temperature of the fuel elements, the margin to nucleate boiling margin and
the graphite temperature.
A series of programs has been devised for RBMK reactors which allow
prompt calculations by the plant computers to ensure thermal stability with
continuous refuelling and with the valves at the channel inlets in any
position. This makes it possible to -determine the thermal parameters of the
reactor for any channel flow frequency and for any type of control (on the
basis of outlet steam quality or the critical power margin) and for any degree
of pre-throttling of the core.
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To determine the power
relies on physical monitorings
and radial neutron flux. In
readings, the plant computer
density fields through the core, the plant
based on in-core measurement of the vertical
addition to the physical monitoring system
also receives data characterizing the core
composition, the energy output of each fuel channel, position of the control
rods, distribution of water flow through the core channels, and readings from
the sensors indicating coolant pressure and temperature. The PRIZMA program
calculations carried out by the computer periodically give the operator a
digital printout of core configurations, indicating the type of core loading,
the position of the control rods, the arrangement of in-core sensors, the
power distribution, the critical power margins and margins for the maximum
permissible thermal loads on the fuel elements for each fuel channel in the
reactor. The plant computer also calculates the overall thermal power of the
reactor, the distribution of steam-water mixture flow through the separators,
the integral power output, the outlet steam quality from each fuel channel and
various other parameters needed to monitor and control the reactor plant.
Experience in operating actual RBMK reactors shows that with the
existing means of monitoring and controlling these reactors there is no
difficulty in maintaining the temperatures of the fuel and graphite and the
critical heat margins at the permissible level.
1.4. Safety systems (Figs 2 and 3)
1.4.1. Protective safety systems
The emergency core cooling system (ECCS) is a protective safety system
designed to draw off the residual heat from the core by feeding an appropriate
volume of water into the reactor channels in the event of accidents which
damage the main core cooling system. Associated with such accidents are
ruptures in the large-diameter MFCC pipelines, as well as ruptures in the
steam pipes and in the feedwater pipelines.
? The system for preventing excess pressure in the main coolant circuit
is designed to ensure an acceptable pressure level in the circuit by drawing
off steam into a pressure suppression pool where it will condense.
The system for protecting the reactor space is designed to ensure that
acceptable pressure within the reactor space is not exceeded in an emergency
situation involving the rupture of one fuel channel; it does this by
transferring the steam and gas mixture from the reactor space into the steam
and gas disposal compartment of the pressure suppression pool and later into
the pressure suppression pool itself with simultaneous suppression of the
chain reaction by the emergency protection system. The ECCS and the reactor
space cooling system can be used to introduce appropriate neutron absorbers
(boron salt and 3He).
1.4.2. Localizing (confining) safety systems
The accident localization system as used on the fourth unit of the
Chernobyl' nuclear power station is designed to localize and contain
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radioactive emissions in accidents involving loss of integrity in any of the
'pipes of the reactor's coolant circuit, with the exception of the pipework of
the steam-water communication lines, the upper parts of the fuel channels and
that part of the downcomers which is situated in the drum separator
compartment, and the pipework for the steam and gas discharges from the
reactor space.
The main component of the localization system is a system of leaktight
enclosures, including the following compartments within the reactor
compartment:
Reinforced leaktight compartments, distributed symmetrically in
relation to the reactor axis and designed for an excess pressure
of 0.45 MPa;
Compartments of the distributing group headers and lower water
communication lines; these compartments are designed in line with
the strength of the elements used in the reactor construction, not
to permit a rise of over 0.08 MPa in the excess pressure level and
are calculated to this magnitude.
The compartments containing the reinforced leaktight compartments and
steam distribution corridor are connected to the water volume of the bubble
condenser by steam discharge channels.
The system of cut-off and sealing devices is designed to ensure
leaktightness in the accident localization area by cutting off the pipelines
linking the sealed and unsealed compartments.
The bubble condenser is designed to condense the steam formed:
In the course of an accident involving loss of integrity of the
reactor circuit;
Through operation of the main safety valves;
By flows through the main safety valves during
1.4.3. Safety-assurance systems (service safety systems)_
Electricity supply to the plant I
normal operation.
The users of electricity at the power plant are divided into three
groups, according to the degree of reliability of supply required:
(1) Those unable to tolerate a break in supply lasting from fractions
?of a second to several seconds under any circumstances, including
complete loss of alternating-current voltage from the plant's own
working and stand-by transformers, and requiring an assured supply
after the reactor's emergency protection system has come into
operation;
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(2) Those which, under the same conditions, can tolerate :a break in
supply lasting from tens of seconds to tens of minutes,, and which
require an assured supply after the reactor's emergency protection
system has come into operation;
(3) Those which do not require a supply in the event of loss of
voltage from the plant's own working and stand-by transformers,
and which, with the unit operating normally, will tolerate a break
in supply during the time taken to transfer from the working
transformer to the plant's own stand-by transformer.
1.4.4. Control safety systems
The control safety systems are designed to automatically bring into
operation the protection, localizing and safety assurance systems and to
monitor their functioning.
1.4.5. Radiation control system
The power plant's radiation control system, which forms an integral
part (i.e. a subsystem) of its automated control system, is designed to
collect, process and display data relating to the radiation situation within
the plant premises and in the external environment, the condition of equipment
and circuits and staff radiation exposure, in accordance with the standards
and legislation, in force.
1.4.6. Power plant control points
There are two levels of control at the plant, namely station level and
unit level.
All systems related to power plant safety are controlled at unit level.
1.5. Description of the site of the Chernobyl' nuclear power station and of
the surrounding region
1.5.1. Description of the region
The Chernobyl' nuclear power station is situated in the eastern part of
a large region, known as the Byelorussian-Ukrainian Woodlands, beside the
,River Pripyat', which flows into the Dnepr. The region is characterized by a
relatively flat landscape with very minor slopes down to the river or its
tributaries.
The total length of the Pripyat' before it flows into the Dnepr is
748 km, and its catchment area at the point where it passes the power plant is
106 000 km2. The river is 200-300 in wide, with an average flow rate of
0.4-0.5 m/sec. The long-time average volume flow is 400 m3/sec.
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The water-bearing horizon used for the above region's drinking water
supply lies at a depth of 10-15 m in relation to the present level of the
Pripyat' and is separated from the Quaternary deposits by relatively
impermeable argillaceous marls.
The Byelorussian-Ukrainian Woodland region is on the whole
characterized by a low population density (up to the start of construction
work on the Chernobyl' power plant the average population density of the
region was approximately 70 inhabitants per km2).
At the beginning of 1986 the total population within a region of
30 kilometre radius around the power plant was approximately 100 000, 49 000
of whom lived in the town of Pripyat', situated to the west of the plant's
three-kilometre safety zone, and 12 500 of whom lived in the town of
Chernobyl', the regional centre, situated 15 km to the south east of the plant.
1.5.2. Description of the power plant site and its buildings
The first stage of the Chernobyl.' power plant, two units with RBMK-1000
reactors was constructed between 1970 and 1977. Work on the two power units
comprising the second construction stage was completed on the, same site in
late 1983.
In 1981 work was begun on the construction of two more power units
using the same reactors (the third construction stage) at a distance of 1.5 km
to the south-east of the existing site. ,
To the south east of the power plant site and directly within the
Pripyat' valley, a 22 km2 cooling water pond was constructed to provide
cooling water for the turbine condensers and the other heat exchangers of the
first four :: units. The normal breast-wall level of the water in the cooling
gond is taken' to be 3.5 m below the design level of the power plant site.
Under the third construction stage, two powerful water-cooling towers
(each with a hydraulic capacity of 100 000 m3/h) are being built; these will
be capable of functioning in parallel with the cooling pond.
The area reserved for the construction base and warehouse facilities is
situated tio, the west and north of the site of the first and second stages.
1.5.3. Information on the number of staff at the power plant site at the 1
of the accident
On the night of 25-26 April 1986 there were 176 duty operational staff
and workers from different departments and maintenance services on the site of
the' first and second construction stages.
In addition to this there were 268 builders and assemblers working on
the night shift on the site of the third construction stage.
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1.5.4. Information on equipment situated on-site and previously in operation
in the complex containing the damaged reactor, and on equipment used in
bringing the accident under control
Construction of the Chernobyl' nuclear power station is being carried
1 out in stages, each comprising two power units with common on-site special
water purification systems and auxiliary facilities, among which are:
i
?\ - A storage facility for liquid and solid radioactive wastes;
- Open distributive systems;
i
- Gas supply unit;
Stand-by diesel power plants;
- Hydraulic and other facilities.
The liquid radioactive waste storage facility, built as part of
construction stage two, is intended for the receipt and temporary storage of
the liquid radioactive wastes arising from the operation of the third and
fourth units, and also to receive water from washing operations and to return
it for processing. The liquid radioactive wastes are channelled from the main
vessel through pipes laid along the lower deck of the pipe bridge, while the
1 solid radioactvie wastes reach the storage facility through the upper corridor
of the pipe bridge in electric trolley-cars.
The nitrogen-oxygen station is designed to supply the needs of the
; plant's third and fourth units.
The gas supply unit comprises a compressor unit, electrolysis unit and
7 helium and argon containers; its purpose is to supply the plant's third and
fourth units with compressed air, hydrogen, helium and argon. Receptacles for
storing the nitrogen and hydrogen are situated in the open.
The stand-by diesel power plant is an independent emergency source of
electricity to supply those systems which are important to the safety of each
unit. Each stand-by diesel power plant of the third and fourth units is
equipped with three diesel generators having a unit output of 5.5 ?W. These
plants are served by an intermediate and a base diesel fuel depot, fuel
transfer pumps and emergency fuel and oil discharge tanks.
The service water for the third and fourth units is supplied by the
cooling water pond.
The water for the circulation pumps, which serves both the third and
fourth units, enters the pressure basin and from there flows by gravity to the
turbine condensers.
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In the case of those users requiring an uninterrupted supply of service
water, this is provided for by separate pumping stations for the third and
fourth units. A stand-by power supply from the diesel generators is available
to these pumping stations.
On 25 April 1986, all four units of the first and second construction
stages were in operation, as were all auxiliary systems and on-site facilities
associated with their normal operation.
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2. CHRONOLOGICAL ACCOUNT OF HOW THE ACCIDENT EVOLVED
The fourth unit of the Chernobyl' nuclear power plant went into
operation in December 1983. At the time when the reactor was tot be shut down
for intermediate maintenance, planned for 25 April 1986, the core contained
1659 fuel assemblies with an average burnup of 10.3 MWd/kg, one additional
absprber and one unloaded channel. Most of the fuel assemblies (75%) were
first load bundles with a burnup of 12 to 15 MWdikg.
Before shutdown, tests were to be carried out on turbogenerator No. 8
in a regime whereby the turbine would be supplying plant power requirements
during the run down. The purpose of these experiments was to test the
possibility of utilizing the mechanical energy of the rotor in a turbo-
generator cut off from the steam supply to sustain the unit's own power
requirements during a power failure. This regime is in fact used in one
sub-system of the reactor's fast-acting emergency core cooling system (ECCS).
If carried out in an appropriate way with the requisite additional, safety
measures, such an experiment would not be forbidden on an operating power
plant.
Similar tests had already been carried out at the Chernobyl' plant. At
that time it had been found that the voltage on the generator busbars falls
off long before the mechanical energy of the rotor is expended during the run-
down. In the tests planned for 25 April 1986 the experimenters intended to
use a special generator magnetic field regulator to eliminate this problem.
However, the "Working Programme for Experiments on Turbogenerator No. 8 of the
Chernobyl' Nuclear Power Plant", in accordance with which these tests were to
be performed, was not properly prepared and had not received the requisite
approval.
The quality of the programme was poor and the section on safety
measures was drafted in a purely formal way. (The safety section said merely
that all switching operations carried out during the experiments were to have
the permission of the plant shift foreman, that in the event ?of an emergency
the staff were to act in accordance with plant instructions and that before
the experiments were started the officer in charge - an electrical engineer,
incidentally, who was not a specialist in reactor plants - would advise the
security officer on duty accordingly.) Apart froth the fact that the programme
made essentially no provision for additional safety measures,, it called for
shutting off the reactor's emergency core cooling system. This meant that
during the whole test period, i.e. about four hours, the safety of the reactor
would be substantially reduced.
Because the question of safety in these experiments had not received
the necessary attention, the staff involved were not adequately prepared for
the tests and were not aware of the possible dangers. Moreover, as we shall
see in what follows, the staff departed from the programme and thereby created
the conditions for the emergency situation.
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On 25 April at exactly 1:00 hours the staff began to reduce the reactor
power (up to then the unit had been operating at rated parameters) and at
13:05 hours turbogenerator No. 7 was switched off with the reactor at
1600 MW(th). The electric power required for the unit's own needs (four main
circulation pumps, two electrical feed pumps and other equipment) was switched
to the busbars of turbogenerator No. 8.
At 14:00 hours the reactor's emergency core
disconnected from the multiple forced circulation circuit
with the experimental programme. However, because
requirements the removal of the unit from operation was
unit then continued to operate with the emergency cooling
in violation of the operating rules.
cooling system was
(MFCC) in accordance
of control room
delayed. Thus, the
system switched off,
At 23:10 hours, the power reduction was resumed. Under the test
programme, the rundown of the generator with simultaneous provision of unit
power requirements was to be carried out at a reactor power of
700-1000 MW(th). However, when the local automatic regulation system was shut
off, which under the operating rules is supposed to be done at low power, the
operator was unable to eliminate the resultant unbalance in the measuring part
of the automatic regulator quickly enough. As a result of this, the power
fell below 30 MW(th). Only at 1:00 on 26 April did the operator succeed in
stabilizing it at 200 MW(th). Since the "poisoning" of the reactor was
continuing at the same time, a further increase in power was hindered by the
small excess reactivity available, which at that moment was substantially
below what the regulations called for.
Even so, it was decided to conduct the experiments. At 1:03 and at
1:07 one additional main circulation pump was switched in from either side to
join the six pumps already operating, so that when the experiment was
finished - during which four main circulation pumps were to be operating
through the rundown - four pumps would remain available on the MFCC for safe
cooling of the reactor core.
Since the reactor power, and consequently the hydraulic resistance of
the core and the MFCC were substantially lower than the planned level and
since all eight main circulation pumps were in operation, the total coolant
flow rate through the reactor rose to (56 000-58 000 m3/h, and at some
individual pumps to 8000 m3/h, which meant a violation of the operating
rules. An operating regime of this kind is forbidden because of the danger of
pump breakdown and the possibility of vibrations arising in the main coolant
pipes owing to cavity formation. The switching in of the additional main
circulation pumps and the resulting increase in water flow through the reactor
brought about a reduction of steam formation, a fall in steam pressure in the
drum separators, and changes in other reactor parameters. The operators
attempted manually to sustain the main parameters of the system - steam
pressure and the water level in the drum separators - but they did not
entirely succeed in doing\ so. At this stage they saw the steam pressure in
the drum separators sag by 0.5-0.6 MPa and the water level drop below the
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emergency mark. In order to avoid shutting down the reactor in such
conditions, the staff blocked the emergency protection signals relating to
these parameters.
At the same time, the reactivity continued to drop slowly. At 1:22:30,
the operator saw from a printout of the fast reactivity evaluation program
that the available excess reactivity had reached a level requiring immediate
shutdown of the reactor. Nevertheless, the staff were not stopped by this and
began with the experiments.
At 1:23:04, the emergency regulating valves of turbogenerator No. 8
shut. The reactor continued to operate at a power of about 200 t4W(th). The
available emergency protection from the closing of the emergency regulating
valves on two turbogenerators (turbogenerator No. 7 had been shut off on
25 April) was blocked so that it would be possible to repeat the experiment if
the first attempt proved unsuccessful. This meant a further departure from
the experimental programme, which did not call for blocking the reactor's
emergency protection with the switching off of two turbogenerators.
Shortly after the beginning of the experiment the reactor power began
to rise slowly.
? At 1:23:40, the unit shift foreman gave the order to press button Az-5,
which would send all control and scram rods into the core. The rods fell, but
? after a few seconds a number of shocks were felt and the operator saw that the
absorber rods had halted without plunging fully to the lower stops. He then
cut off the current to the sleeves of the servo drives so that the rods would
fall into the core under their own weight.
According to observers outside unit 4, at about 1:24 there occurred two
explosions one after the other; burning lumps of material and sparks shot into
the air above the reactor, some of which fell onto the roof of the machine
room and started a fire.
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3. ANALYSIS OF THE ACCIDENT USING A MATHEMATICAL MODEL 1
The "Skala" centralized control system of the RBMK-1000 reactor has a (
program for diagnostic parameter recording (DPRP) under which several hundred i
i
analog and discrete parameters are periodically examined and stored in 1
accordance with a specified cycle (minimum cycle time 1 second).
In connection with the experiments, only those parameters were recorded
with great frequency which were important for an analysis of the experimental
results. Therefore, in trying to reconstruct the course of the accident, we
used a mathematical model incorporating not only the DPRP print-out but also
instrument readings and the results of questioning of the staff.
To perform a rapid analysis of different variants and versions of the
accident situation under consideration, we used an integral mathematical
computer model of the RBMK-1000 unit in real time. The dependences of
reactivity on steam content and on absorber rod movements were determined from
calculations based on three-dimensional neutron physics dispersion models.
In reconstructing the course of the accident, it was particularly
important to be sure that the mathematical model correctly described the
behaviour of the reactor and other equipment and systems in precisely those
conditions which prevailed just before the destruction of the unit. As we
have already noted in the previous chapter, the reactor was operating unstably
after 1:00 hours on 26 April 1986 and the operators were almost continualy
introducing new "perturbations" into the controlled system in order to
stabilize its parameters. This has made it possible, for a fairly long time
interval involving various influences on the reactor, to compare factual data
established fairly reliably by the recording systems with the data obtained
through numerical modelling. The results of this comparison have proved to
be highly satisfactory, which suggests that the mathematical model
satisfactorily reproduced the actual plant.
In order to get as clear an idea as possible of the influence of
preceding events on the development of the accident, we analysed data
beginning at 1:19:00, i.e. 4 minutes before the beginning of the turbo-
generator run down experiment (Fig. 4). This movement is convenient in the
sense that the operator was then starting one of the operations involved in
the drum-separator make up (the second since 1:00 hours) which produced
powerful perturbations in the controlled system. At this moment the DPRP
recorded the position of the rods of all three automatic regulators - in other
words, the initial conditions of the calculation were very clearly established.
The operator began the drum separator make up in Order to prevent a
radical drop in the water level of the separators. After 30 seconds he
succeeded in maintaining the level by increasing the input flow of make up
water by a factor of more than three. It would seem that the operator had
decided not only to maintain the water level but to raise it. For that reason
the water flow continued to increase and after about a minute was already four
times the initial value.
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As soon as the colder water from the drum separators reached the core,
steam generation was substantially reduced, and this in turn reduced the
volumetric steam quality, raising the automatic regulator rods. Within about
30 seconds the rods rose to the upper stops and the operator had to "help"
them with the manual control rods, thereby reducing the available excess
reactivity (this operation was not recorded in the daily operating log, but
without it the operator could not possibly have maintained the power at
200 MW). By moving the manual rods upwards the operator brought about an
over-compensation and one of the groups of automatic regulator rods dropped
1.8 metres.
The reduction in steam generation brought about a small drop in
pressure. Within about a minute, at 1:19:58, the fast-acting steam dump
system was closed off, through which excess steam had been slipping to the
condenser. This slowed down the rate of pressure drop a little. Even so, up
to the beginning of the experiment the pressure continued to fall off slowly.
During this period it changed by more than 0.5 MPa. At 1:22:30 the "Skala"
centralized control system provided a print-out of the actual power density
fields and of all regulatory rod positions. It was for this instant in time
that we attempted to correlate the calculated and recorded neutron fields.
The overall neutron field characteristic at this moment can be
described as follows: in the radial-azimuthal direction it showed for all
practical purposes a smooth convex shape, but in the vertical direction the
curve was double-humped, on average, with a greater release of energy in the
upper part of the core. A neutron field distribution of this kind would be
completely natural for the state prevailing in the reactor at that moment: a
burnt-out core, practically all regulating rods up, volumetric steam quality
in the upper part of the core much more than lower down, and greater 135Xe
poisoning in the central parts of the reactor than on the periphery.
At 1:22:30 the excess reactivity was only 6-8 rods, in other words not
more than half of the minimum permissible value laid down in the operating
regulations. The reactor was in an unusual and impermissible state, and to
assess the subsequent course of events it was extremely important to determine
the differential rod worths of the control rods and the scram rods for real
neutron fields and real core multiplication characteristics. Numerical
analysis showed that the error in determining the control rod worths was
extremely sensitive to the error in re-establishing the vertical power density
field. Add to this the fact that at such low power levels (approximately
6-7%) the relative error in measuring the field is much greater than in rated
power conditions, then it becomes clear that a vast number of calculational
variants will have to be analysed before one can be confident of the rightness
or wrongness of any given version.
At 1:23 hours the reactor parameters were closer to stable than at any
other time in the interval we are considering, and the experiments began. A
minute before this the operator had abruptly reduced the flow of make up
water, and this increased the water temperature at the reactor inlet within a
time equivalent to that required for the coolant to pass from the drum
separators to the reactor. At 1:23:04 the operator closed the emergency
regulating valves of turbogenerator No. 8 and the turbogenerator rundown
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began. Because .of. thp reduced flow of steam from. the...drum ,.separators, the
.steariv-.pr6ssure began_torjse slowly .(on average at a rate of 6 kPa/s).... The
-
A..c4411 flow' of water through the: reactor began to fall off owing to the fact
that le4kr of the eight'Main-cf.rculatien pumps Were working off the "running
,? ?
down" tUrbegeherater-.
The inerease in Steam pressure on the one hand and the reduced flow of
water through the reactor together with the reduced input of make up water to
' the drum separators, on the other, hand, are competing factors in determining
the volumetric steam quality and hence the power of the reactor. A point that
deserves particular stress is that in the state the reactor had now reached a
small change in power would mean that the volumetric steam quality - which has
a direct effect on reactivity - would increase much more than at nominal
power. The competition between these factors led in the final analysis to a
power rise, and this was the circumstance which triggered the pressing of
button AZ-5.
Button AZ-5 was pushed at 1:23;40 and the insertion of the scram rods
began. At this time the automatic regulator rods, partially compensating for
the previous power rise, were already in the lower part of the core, but the
fact that the staff were operating with an impermissibly small excess
reactivity meant that virtually all other absorber rods were in the upper part
of the core.
In the conditions that had now arisen, the violations committed by the
staff had seriously reduced the effectiveness of the emergency protection
system. . The overall positive reactivity appearing in the core began to
increase. Within three seconds the power rose above 530 MW, and the total
period of the excursion was much less than 20 seconds. The positive void
coefficient of reactivity worsened the situation. The only thing that
partially compensated for the reactivity inserted at this time was the Doppler
effect,
? The continuing reduction of water flow through the fuel channels as the
power rose led to an intensive steam formation and then to nucleate boiling,
over-heating of the fuel, destruction of the fuel, a rapid surge of coolant
boiling with particles of destroyed fuel entering the coolant, a rapid and
abrupt, increase of pressure in the fuel channels, destruction of the fuel
channels, and finally an explosion which destroyed the reactor and part of the
building and released radioactive fission products to the environment.
ln the mathematical model, the destruction of the fuel was simulated by
an abrupt increase in effective heat exchange surface when the power density
in the fuel exceeded 300 cal/g. It was precisely at that moment that the
pressure in the core had risen to the point where an abrupt reduction of water
-supply from the main circulation pumps occurred (the cheek valves were
closed). This is quite plain from the results obtained with the mathematical
model and from the results and measurements recorded by the DPRP. Only the
rupture of the fuel channels partially restored the flow from the main
circulation pumps; however, the water from the pumps was at this stage no
longer directed into intact channels but into the reactor space.
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The steam formation and rapid rise of temperature in the core created
appropriate conditions for a steam-zirconium reaction and other exothermal
chemical reactions. Witnesses observed these reactions in the form of a
fireworks display of glowing particles and fragments, escaping from the units.
As a result of these reactions, a mixture of gases was formed
containing hydrogen and carbon monoxide, which then led to a thermal explosion
upon mixing with the oxygen of the air. This mixing became possible after the
reactor space had been vented and destroyed.
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4. ' CAUSES OF THE ACCIDENT
As shown by the analysis presented above, the accident in the fourth
unit of the Chernobyl' nuclear Power plant belongs to the category of
accidents associated with the introduction of excess reactivity. The design
of the reactor facility provided for protection against this type of accident
with allowance for the physical characteristics of the reactor, including a
positive steam void coefficient of reactivity.
-The technical means of 'protection include systems for controlling the
reactor and protecting it against a power overshoot, for reducing the
examination period, and for self-shielding and protection against
malfunctioning in switching operations involving the equipment and systems of
the power-generating unit and the emergency core cooling system.
Apart from the technical means of protection, there are also strict
rules and instructions for carrying out technological processer at a nuclear
power plant, specified in regulations for the operation of. each power-
generating unit. Among the most, important regulations are stipulations
referring to the inadmissibility of reducing 'theoperational excess reactivity
(reactivity margin) to fewer' than 30 -rods.
in the process of preparing for and conducting the turbogenerator
tests, in which the turbine was to supply the unit's requirements during the
run-down:, the staff switched off a number of important protection systems and
violated the most important provisions of the operating regulations' for Safe
management of technological process.
' The table below lists the most dangerous violations of the operating
rules committed by the staff of the fourth unit of the Chernobyl' nuclear
power plant.
No. Violation Motivation
1 2 3
1. Reducing the operational
reactivity margin sub-
stantially below the
permissible value
2. Power dip well below the
level provided for by ,
the test programme
3. Connecting of all the
main circulation pumps
to the reactor, with
individual pump dis-
charges exceeding the
levels specified in the
regulations -
Attempt to emerge
from,"iodine well"
Operator error in
switching off local
automatic control
Meeting the
requirements of
the tests
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Consequences
4 '
Emergency protection
system of reactor was
ineffective
Reactor proved to be in
a condition difficult to
control
Coolant temperature in
the multiple forced
circulation circuit
approached saturation
temperature
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4. Blocking of reactor
protection system relying
on shutdown signal from
two turbogenerators
S. Blocking of protection
systems relying on water
level and steam pressure
in the drum-separator
6. Switching off of the
protection system for the
design-basis accident
(switching off of the
emergency core cooling
system)
Intention, if
necessary, of
repeating the
experiment with
turbogenerators
switched off
Effort to perform
tests despite
unstable reactor
operation
Wish to avoid
spurious triggering
of the emergency
core cooling system
while the experiment
was going on
Loss of possibility of
automatic shutdown of
the reactor
Reactor protection system
based on heat parameters
was completely cut off
Loss of the possibility
of reducing the scale of
the accident
The chief motive of the staff was to complete the tests as expeditiously
as possible. The failure to adhere to instructions in preparing for and
carrying out the tests, the non-compliance with the testing programme itself
and the carelessness in handling the reactor facility are evidence that the
staff was insufficiently familiar with the special features ,of the
technological processes in a nuclear reactor and also that they had lost any
feeling for the hazards involved.
The designers of the reactor facility did not provide for protective
safety systems capble of preventing an accident in the combination of
circumstances prevailing in unit 4, namely the deliberate switching off of
technical protection systems coupled with violations of the operating
regulations, since they considered such a conjunction of events to be
impossible.
Thus, the prime cause of the accident was an extremely improbable
combination of violations of/ instructions and operating rules committed by the
staff of the unit.
The accident assumed
taken by the staff into a
coefficient of reactivity
excursion.
catastrophic proportions because the re,6tor was
non-regulation state in which the positive void
was able substantially to enhance the power
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5. pR1OR1TY MEASURES FOR IMPROV1NG-THE SAFETY OF NUCLEAR POWER PLANTS WITH
RBMK REACTORS
It has been decided, for existing nuclear power plants with RBMK
reactors* to alter the limit stop switches of the control rods in such a way
- that, in the ,extreme position, all the rods are inserted in the core to a
. depth of 1.2 m. This measure will increase the speed of effective protection
and eliminate the possibility of -a continuing increase in the multiplying
characteristics of the core in its lower part as the rods move down from the
upper stops. At the same time, the number of absorber-type control rods
constantly.present in the core will be increased to 70-80, thereby reducing
void coefficient to a permissible value. This is a temporary measure
which'will. be replaced later on by a conversion of RBMK reactors to fuel with
an initial enrichment' of 2.4% and by the insertion of additional absorbers in
the core, to ensure that a positive overshoot of reactivity does. not-exceed
1,0 for any change, in coolant density..
A number of additional indicators of the cavitation margin of the main
circulating pumps are being installed, and also a system for.- automatic
calculation of reactivity with an emergency shutdown signal when the excess
reactivity falls below ,a specified level. These measures will have?a somewhat
adverse effect on the economic parameters of nuclear power plants with RBMK
reactors but they will guarantee safe operation.
In addition to the technical measures, organizational steps are being
taken to reinforce technological discipline and to improve the quality of
operations.
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6. CONTAINMENT OF THE ACCIDENT AND ALLEVIATION OF ITS CONSEQUENCES
[
6.1. Fighting the fire at the nuclear power station
i
The most important task after the accident at the reactor was to fight
the fire. Fire had broken out in over 30-places as a result of the explosions
in the reactor which had ejected fragments of its core, heated to high
temperatures, onto the roofs above, severaF'areas hOuaint the reactor section,
the de-aeration stages and the machine hall. Because' of damage to some ,oil
pipes, electric cable short circuits and the intense heat radiation from the
? reactor, focuses of fire formed in the machine' hall Over turbogenerator
number 7, in the reactor hall and in the adjoining,, partiallyde?troyed
? buildings. At 01.30, the firemen on duty 'with the subsection ofr'the fire
division responsible for the power station set out from the towns of'Pripyats
and Chernobyl' to the scene of the accident. In view of the immediate threat
that the fire would spread along the top of the machine hall tb the' adjoining
third unit, and as it was rapidly increasing in strength, the, first set of
measures taken was directed towards putting out the fire in th-is critical
area. It,was therefore decided that the fires inside the buildings' 'should be
put out with fire extinguishers and .the fire,hydrants installed inside. The
main focuses had been overcome by 02.10 for the machine' hall'roof and by 02.30
for the roof of the reactor section. ''The fire was out by 05.00.
6.2. Evaluation of the state of the fuel after the accident
The accident partially destroyed the reactor core and completely
destroyed its cooling system. This being the case, conditions in the reactor
vault were determined from:
- Residual heat released by the fuel as
fission products;
a result of the decay of
- Heat production from various chemical reactions in the reactor
vault (hydrogen burning, oxidation of graphite and zirconium and
so on);
- Heat removal from the reactor vault through cooling by atmospheric
air through openings in the previously hermetic compartments
around the core.
In order to prevent the accident from spreading and limit its
after-effects, significant efforts were directed during the very first hours
after the accident towards evaluating the condition of the fuel and any
possible change in that condition with the passage of time. To this end, it
was necessary to carry out investigations as follows:
- To evaluate the possible scale of melting (as a result of residual
heat production) of the fuel in the reactor vault;
- To study the interaction between the melted fuel and the reactor
structural materials and vault (metals, concrete and so on);
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- To evaluate the poElsibility that the reactor structural materials
and vault might melt because of the heat released by the fuel.
In'the first place, calculations were carried out to evaluate the state
of the fuel in the reactor vault, taking into account the leakage of fission
products as a function of the time elapsed since the accident.
An analysis of the dynamics of fission product leakage from the reactor
in the first few days after the accident indicated that the change in fuel
temperature with the passage of time was not monotonic. It could be assumed
that there had been several stages in the fuel temperature regime. At the
moment of the explosion, the fuel had heated up. An estimate of temperature
based on the relative leakage of the iodine nuclides (that fraction of the
total isotope content of the fuel which was escaping at any one moment)
indicated that the effective temperature of the fuel remaining in the reactor
building was 1600-1800 K after the explosion. During the next few tens of
minutes, the fuel temperature decreased through heat transfer to the graphite
structure and structural materials. There was therefore a corresponding
reduction in the leakage of volatile fission products from the fuel.
It was considered that the quantity of fission products ejected from
the reactor vault was fundamentally determined during this period by graphite
burning and the related migration processes of the finely dispersed fuel and
fission products embedded in the graphite as a result of the explosion in the
reactor. Subsequently, the fuel temperature began to increase because of
residual heat production. As a result, the leakage of volatile radionuclides
from the fuel increased (inert gases, iodine, tellurium and caesium). When
the fuel temperature had increased further, other, non-volatile radionuclides
began to escape. By 4-5 May the effective temperature of the fuel still in
the reactor had stabilized and then began to decrease.
The results from the numerical calculation of the condition of the fuel
are shown in Fig. 5. The figure shows the residual radionuclide content of
the fuel, and also the variation in the fuel temperature taking into account
the leakage of fission products as a function of time elapsed since the
accident.
The calculations show:
- That the maximum fuel temperature could not have reached the
melting point of the fuel;
- That the fission products were coming to the fuel surface in
batches, which could lead to only local overheating at the
fuel-cladding interface.
The fission products leaving the fuel settled on the structural and
other materials surrounding the reactor in the unit, according to their
temperatures of condensation and precipitation. Virtually all the krypton and
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xenon radionuclides left the unit, some of the volatile fission products
(iodine, caesium) did so, and practically all the rest stayed within the
reactor building. The energy from the fission products was thus dispersed
throughout the reactor unit.
These factors indicate that melting of the materials surrounding the
fuel and movement of the fuel were unlikely.
_ 6.3. Limiting the consequences of the accident in the core
The potential danger that some melted fuel would concentrate, creating
conditions in which a critical mass might be reached and a spontaneous chain
reaction occur, made it necessary to take appropriate precautions. In
addition, /the damaged reactor was releasing significant amounts of
radioactivity into the environment.
Immediately after, the accident, an attempt was made to reduce the
temperature in the reactor vault and to prevent the graphite structure
igniting by using the /emergency auxiliary feed pumps to supply water to the
core space. This attempt proved ineffective.
One of two decisions had to be taken immediately:
- To contain the accident at source by covering the reactor shaft
With heat/absorbent and filtering materials;
- To allowithe combustion processes in the reactor shaft to come to
an end of their own accord.
The 'first line of action was chosen, as the second carried within
itself the danger that a significant area would suffer radioactive
contamination and the health of the inhabitants of major cities might be
threatened.
A group of specialists began to cover the damaged reactor by dropping
compounds of boron, dolomite, sand, clay and lead from military helicoptors.
About 5000 t in all were dropped between 27 April and 10 May, mostly between
28 April and 2 May. As a result, the reactor was covered with a friable layer
of material which strongly absorbed aerosol particles. By 6 May, the release
of radioactivity had ceased to be a major factor, having decreased to a few
hundred Ci, and fell to a few tens of Ci per day by the end of the month.
The problem of reducing the fuel temperature was solved at the same
time. To bring down the temperature and reduce oxygen concentration, nitrogen
was pumped under pressure from the compressor station into the space beneath
the reactor vault. By 6 May, the temperature increase in the reactor vault
had ceased, and had begun to reverse itself with the formation of a stable
convective flow of air through the core into the open atmosphere.
As a form of double insurance- against the extremely low risk (although
it was a possibility in the first few days after the accident) of the lower
levels of the structure being destroyed, the decision was taken to construct,
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'As a Matter of urgency, an artificial heat-removal horizon beneath the
loundationi,of`the building. This took 'theform of a flat heat-exchanger on :a
concrete slab. This 'had been done by the 'end of June.
'Experience has shown that the decisions which were taken were basically
correct.
A significant degree of stabilization has taken place since the end of
May. The damaged parts of the reactor building are stable, and the radiation
situation is improving now that the short-lived isotopes have decayed. The
exposure dose rate is in the single rOntgens per hour range in the areas
adjoining the reactor, the machine hall and control and protection areas. Any
uptake of radioactivity from the unit into the atmosphere is basically caused
by wind removing aerosols. The activity of the releases does not exceed some
tens of curies per day. The temperature regime in the reactor vault is
stable. The maximum temperatures of the various reactor parts are a few
hundreds of degrees centigrade and they have a steady tendency to fall at
about 0.50C per day. The slab at the base of the reactor vault is intact,
and the fuel is mostly (- 96%) localized within the reactor vault, and the
compartments of the steam-water and lower water lines.
6.4. Measures taken at units 1, 2 and 3
After 'the accident in the fourth unit, the following measures were
taken at units 1, 2 and 3:
Units 1 and 2 were shut down at 01.13 and 02.13 respectively on
27 April;
Unit 3, which is closely linked technically with the damaged
fourth unit, but which suffered practically no damage from the
explosion, was shut down at 05.00 on 26 April;
Units 1 to 3 were prepared for a lengthy cold shutdown;
After the accident, the power station equipment was placed in the
cold reserve state.
Units 1 to 3 and the power station equipment are checked by the staff,
on duty. Significant radioactive contamination of the equipment and buildings
of units 1 to 3 of the power station was caused by radioactive materials
coming through the ventilation system, which continued to operate for some
time after the accident. There was a significant degree of radiation in some
parts of the machine hall, which was contaminated through the damaged roof of
the third unit.
The Government Commission ordered that decontamination and other work
should be carried out on the first, second and third units with the aim of
preparing them eventually for startup and operation again.
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Special solutions were used for decontamination. Their composition
depended on the material being cleaned (plastic, steel, concrete, various
coverings) and the type and level of surface contamination. After
decontamination, gamma radiation levels dropped by a factor of 10 to 15. Dose
rates within the first and second units were between 2 and 10 mR/4 in June.
Final decontamination and stabilization of the radiation situation at the
first, second and third units can only take place when decontamination work
has been completed for the rest of the power station area and when the damaged
unit has been entombed.
6.5. Monitoring and diagnosis of the state of the damaged unit
The organization of diagnostic measurements envisaged the resolution of
the following basic problems:
- Establishment of reliable monitoring of fuel displacement;
- Determination of the scale of contamination in the area adjacent
to the nuclear power plant;
- Evaluation of damage and dosimetric survey inside the unit, and
determination of possibilities of work in the surviving premises;
Determination of the distribution of fuel, fission products etc.
in order to work out the basic data for designing structures for
entombment.
Apart from evaluation of the radiation situation in and around the
plant, the priority measurements included monitoring of the condition of the
reactor from the air. Helicopters were used to carry out radiation
measurements, an infrared survey of the damaged reactor building and its
components with a view to measuring the temperature field distribution,
analysis of the chemical composition of gases emitted from the reactor vault
and a number of other measurements. After it had been determined that the
premises and equipment had survived in the lower part of the reactor building,
it became possible to conduct .initial measurements and to install emergency
monitoring instruments. First of all, instruments for measuring neutron flux,
gamma dose rate, temperature and heat flow were installed in the evacuated
pressure suppression pool. Redundancy was provided for the thermometric
instruments. Evaluation of the situation in the pressure suppression pool
showed that there was no imminent danger of melting of structural parts. This
afforded the assurance that work on construction of a protective slab beneath
the unit could be carried out under safe conditions.
lines:
The general measurement strategy was cormulated along the following
Dosimetric and visual survey inside the damaged unit;
Radiometric and visual survey from helicopters;
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- Measurement of the most important parameters (radioactivity,
temperature and air flow) in the surviving structures and
accessible premises.
The main measurement efforts at the initial stage were concentrated on
monitoring any downward displacement of the fuelthat might occur.
Solution of the diagnostic problem was complicated by the following
factors:
- The regular measurement system had broken down completely;
- The outputs of any detectors which might have survived were
inaccessible to the personnel;
- Information on the condition of compartments and rooms, and on the
radiation situation in them, was limited.
At the next stage it was necessary to determine the location in the
building of the fuel ejected from the reactor vault and to evaluate its
temperature and heat removal.
Conventional or dosimetric survey methods were used to deal with this
problem; in addition, some surviving process pipelines were found through
which the measurement probes could be inserted. As a result of these
investigations, the distribution of the fuel inside the building was largely
determined.
The temperature in the compartments under the reactor did not exceed
450C as from June, indicating good heat removal.
The monitoring and diagnostic methods were refined on the basis of the
data obtained.
6.6. Decontamination of the site
At the time of the accident radioactive materials were scattered over
the site and fell on the roof of the turbine hall, the roof of the third unit
and on metal'pipe supports.
The site as a whole as well as the walls and tops of buildings likewise
had substantial contamination as a result of deposition of radioactive
aerosolt and radioactive dust. The contamination of the site was not uniform.
With a view to reducing the spread of radioactive dust from the site,
the roof of the turbine hall and the road shoulders were treated with various
polymerizing solutions in order to immobilize the upper layers of soil and
prevent dust from rising.
The plant area was divided into separate zones with a view to
compehensive decontamination operations. Decontamination is being carried out
in the following sequence:
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- Removal of refuse and contaminated equipment from the site;
- Decontamination of roofs and outer surfaces of buildings;
- Removal of a 5-10 cm layer of soil and its transfer, in containers
to the solid waste repository of the fifth unit;
Laying, if necessary, of concrete slabs on the soil or filling
with clean earth;
- Coating of the slabs
film-forming compounds.
and of the non concrete area with
As a result of the above measures, it has been possible to reduce the
total gamma background in the area of the first unit to 20-30 mR/h. This
residual background is due mainly to external sources (damaged unit),
indicating that the decontamination of the site and buildings has been
sufficiently effective.
6.7. Long-term entombment of the fourth unit
Entombment of the fourth unit should ensure a normal radiation
situation in the surrounding area and in the atmosphere and preclude escape of
radioactivity into the environment.
For purposes of entombment of the unit it is intended to build the
following engineering structures (Figs 6-8):
- Outer protective walls along the perimeter;
- Inner concrete partition walls in the turbine hall between the
third and fourth units, in the "B" block and in the de-aerator
room along the turbine hall and on the side of the debris by the
tank room of the emergency core cooling system;
- A metal partition wall in the turbine hall between the second and
third units;
- A protective roof over the turbine hall.
Furthermore, it is planned to seal off the central hall and other
reactor rooms and to pour concrete over the debris by the tank room of the
emergency core cooling system and over the rooms of the northern main
circulation pumps to isolate the debris and to provide protection against
radioactive radiation from the reactor sector.
The thickness of the protective concrete walls will be 1 m or more,
depending on the design solutions adopted and the radiation situation.
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,Two variants are considered in the ventilation design:
An open system with purification of air by aerosol filters and
release into the atmosphere through the existing stack of the
central ventilation plant;
A closed system :with removal of heat into the heat exchanger
located in the upper part of the space to be ventilated and
maintenance of negative pressure inside the building to be ensured
by pumping of air from the upper part of the space and its
discharge through filters and the stack into the atmosphere.
The above activities are to be carried out in the following sequence:
The surface layer of soil in the area adjacent to the unit is to
be removed to local sites by means of special technology;
2. The area will be covered with concrete and the surface levelled to
ensure the movement of self-propelled cranes and other machinery;
3. The roofs and walls of buildings are to be decontaminated.
At locations where radioactivity is high, special polymer adhesive
pastes of various compositions will be used;
4. After the site has been cleaned and covered with concrete, the
metal frames for the protective walls will be assembled and then
concrete will be provided;
5. As the walls are built, work will proceed with the construction of
the main civil engineering structures which are to ensure complete
emtombment of the fourth unit.
6.8. Decontamination of the 30 km zone and its rehabilitation for economic
use
Significant radioactive contamination of the areas adjacent to the
power plant made it imperative to take a number of extreme decisions involving
the establishment of surveillance zones, evacuation of the population, bans or
restrictions on economic use of land and so on.
It was decided to establish three surveillance zones: a special zone,
a 10 km zone and a 30 km zone. In these zones, strict dosimetric monitoring
of all transport has been organized and decontamination points have been
established. At the zone boundaries there are arrangements for transferring
working personnel from one vehicle to another in order to reduce transmission
of radioactive substances.
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The radiation situation within the 30 km zone will continue to change,
especially in areas with a high gradient of contamination levels. There will
occur a substantial redistribution of radionuclides over the different parts
of the landscape, depending on the characteristics of the topography. The
question of re-establishing of the population can be raised only after the
radiation situation over the whole contaminated area has been stabilized, by
entombment of the fourth ,unit, decontamination of the plant site and
immobilization of radioactivity in locations with a high contamination level.
In June, construction of a complex of hydraulic engineering structures
began with a view to protecting from contamination the ground water and
surface water in the Chernobyl' nuclear power station area. These include:
- A filtration-proof wall in the soil along part of.the perimeter of
the industrial site of the power plant and wells for lowering the
water table;
- A drainage barrier for the cooling pond;
- A drainage cut-off barrier on the right bank of the river Pripyat';
- A drainage interception barrier in the south-western sector of the
power plant;
- Drainage water purification facilities.
On the basis of an assessment of the soil and plant contamination in
the 30 km zone, special agro-engineering and decontamination measures have
been worked out and are now being implemented. Work aimed at restoration of
the contaminated land to economic use has yet started, thanks to these
measures, which include: changes in the conventional systems of soil
treatment in this region, use of special compositions for dust suppression,
modification of harvesting and crop processing methods, and so on.
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7. MONITORING OF ENVIRONMENTAL RADIOACTIVE CONTAMINATION AND HEALTH OF THE
'POPULATION
7.1. Assessment of the quantity, composition and dynamics of the release of
fission products from the damaged reactor
The assessment was based on the results of the following:
- Systematic analyses of the radioisotopic compocition of aerosol
samples collected at points above the damaged unit from
26 April 1986 on;
Airborne gamma survey of the plant area;
- Analysis of fallout samples;
- Systematic
stations.
measurement data from the country's meteorological
The release of radioisotopes from the damaged unit took place over an
extended period of time which can be divided into several stages.
In the first stage there was a release of dispersed fuel from the
damaged reactor. The radioisotopic composition at this point corresponded
roughly to that of the irradiated fuel, but enriched by volatile isotopes of
iodine, tellurium, caesium and inert gases.
In the second stage - from 26 April to 2 May 1986 - the rate of release
from the unit decreased as a result of the measures taken to stop the graphite
burning and to filter the releases. During this period the composition of the
radioisotopes being released Was again similar to that in the fuel. During
this stage finely dispersed fuel was being carried out of the reactor by a
flow of hot air and the graphite combustion products.
The third stage was marked by a sharp increase in the rate of release
of fission products from the unit. In the initial phase of this stage, the
release was composed mainly of volatile components, especially iodine, but
then the radioisotopic composition once more became similar to that of the
irradiated fuel (on 6 May 1986). The reason for this was the heating of the
fuel in the core to a temperature exceeding 1700?C as a result of the
reactor after-heat. The temperature caused the migration of fission products
and the chemical transformation of uranium oxide, which in turn led to an
escape of fission products from the fuel matrix and their relsease in aerosol
form on the graphite combustion products.
The fourth and last stage, which began after 6 May, was characterized
by a rapid drop in releases (Table 1). This was a consequence of the special
measures taken, the formation of more infusible fission product compounds as a
result of their interaction with the materials introduced, and the
stabilization and subsequent lowering of the fuel temperature.
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The radioisotopic composition of the releases is shown in Table 2.
The fission products in the air and fallout samples were in the form of
individual radioisotopes (mainly, volatile ones), , and were .part of the
composition of the fuel particles. Particles (associates) with an elevated
content of individual radioisotopes ,(Cs? Ru ,etc,.,) were identified, these
having formed as a result of the migration of fission products in the fuel and
the filling and structural materials and of sorption on surfaces.
The total release of fission products (excluding radioactive inert
gases) was approximately 50 MCi, or about 3.5% of the total inventory of
radioisotopes in the reactor at the time of the accident. These figures were
calculated on 6 May 1986 and take into account radioactive decay. The release
of radioactive materials virtually ceased on that day.
The composition of the radioisotopes released during the accident
corresponded approximately to that of the fuel of the damaged reactor, the
difference being that the former had a higher content of volatile iodine,
tellurium, caesium and inert gases.
7.2. Monitoring system
When the accident occurred, the official meteorological, radiation and
public health monitoring system began to operate on an emergency footing. As
soon as the scale of the accident became evident, the monitoring system was
widened to bring in additional groups of experts and technicians. In the
first days after the accident, efforts were concentrated on the most urgent
radiation, public health and biomedical monitoring tasks.
During ths period the monitoring system began to be extended to cover
long-term problems also. Among the organizations involved in the
establishment of the system were the State Committee on Hydrometeorology and
Environmental Protection, the Ministries of Health of the USSR and of the
Union Republics, the Academies of Science of the USSR, the Ukrainian SSR and
the Russian SSR, the State Committee on the Utilization of Atomic Energy and
the State Agro-industrial Committee.
The help of specialized medical institutions in Moscow and Kiev was
enlisted to treat those exposed to radiation.
In addition to setting up a monitoring system, programme of
radioecological, biomedical and other scientific studies to evaluate and
predict the effects of the ionizing radiation on man, the flora and fauna was
drawn up and began to be implemented.
The priority objective of the monitoring programme were as follows:
Assessment of the possible internal and external exposures of
plant personnel and the population of Pripyat' and the 30 km zone;
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Assessment of the possible exposure of the population of a number
of areas outside the 30 lesi _zone, the level of radioactive
contamination in which could, have exceeded the permissible limits;
Preparation of recommendations on measures to protect the
population and staff from exposures in excess of the established
limits.
These recommendations included:
Evacuation of the population;
- Restrictions Or a ban ,on the use of food prodUcts
increased amounts of radioactive substances;.
containing
Recommendations on what action people at home and in open places
should take.
In order to solve these priority problems., systematic monitoring was
introduced in respectof the following:
The level of gamma radiation in contaminated areas;
- The concentration of biologically significant radioisotopes in the
air and water of water bodies, particularly those supplying
drinking water;
The degree of radioactive contamination of the soil and vegetation
and its radioisotopic composition;
- The amount of radioactive substances in food products, especially
I in'milk131 "
- Radioactive contamination of working and non-working clothes,
footwear, means of transport etc.;
Build up of radioisotopes in internal organs of people etc..
7.3. Main characteristics of the radioactive contamination of the atmosphere
and ground and possible ecological consequences
The determining factors in the radioactive contamination of the
environment as a result of the. Chernobyl' accident were the dynamics of the
radioactive releases and the Meteorological-conditions.
The radioactively 'contaminated plume moved first to the west and north;
during the.: 2-3 days after the accident - to the north; and, for a few days
from 29 April - to the south. The contaminated air masses then dispersed for
great distances over the Byelorussian, Ukrainian and Russian Soviet Socialist
Republics. The height of the -plume on 27 April exceeded 1200 in, while the
?
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radiation levels in it were 1000 mR/h ,,at -a ..distance of 5-10 km from the
accident site. The plume and the radioactive track which was forming were
regularly surveyed by the airplanes of the State Committee on Hydrometeorology
and Environmental Protection, which were equipped with sampling and gamma
spectrometry equipment and Roentgen , )meters, ,,and by ,the network of
meteorological stations.
Fission and induced activity products
Identified in the air samples.
(239Np
and
134cs)
were
The main zones of ground contamination after Lhe accideni were to the
west, north-west and north-east of the Chernobyl' plant, and subsequently -
and to a lesser extent - to the south. ,Radiation levels near the plant
exceeded 100 mR/h; on the western track the maximum radiation levels 15 days
after the? accident were 5 mR/h at a distance of 50-60 km from the accident
zone (maximum distances), and the same to the north at a distance of
35-40 km. In Kiev the radiation levels at the beginning of May reached
0.5-0.8 mR/h.
In the zone of the radioactive track near the plant, in addition to the
isotopes listed above, plutonium isotopes were identified (their distribution
on the ground was insignificant). In this zone isotope fractionation was
insignificant, but on the remote radioactive track the radioactive products
were considerably enriched by tellurium, iodine and caesium isotopes.
By integrating the contaminated areas it was possible to determine the
total activity of the radioactive fallout outside the plant site. In the
nearby and remote fallout areas on the European territory of the Soviet Union
it amounted to about 3.5% (see subsection 7.1) of ?the total activity of the
fission and activation products accumulated in the reactor (about 1.5-2% on
the nearby trail).
Summing the activity (which was determined by taking ground samples) of
the radioisotopic fallout on the nearby track yielded an approximate value of
between 0.8% and 1.9%.
The plutonium isotope contamination levels in the zones mentioned above
were not the determining as regards decontamination work and decisions of an
economic nature.
Information on the radioactive contamination of rivers and water bodies
was obtained by regularly analysing water samples ,from the Rivers Pripyat',
Irpen', Teterev, Desna and the Dnieper water intake. From 26 April 1986 on
water samples were collected from the whole water area of the Kiev reservoir.
The highest 1311 concentrations were observed in the Kiev reservoir on
3 May 1986, the figure being 3.10-8 Ci/L. It should be noted_ that the
spatial distribution of radioisotopes in the aquatic environment, was very
uneven.
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From the first days of the accident, monitoring of the radioisotopic
content Of sediments at the bottom of water bodies both inside and outside the
30 km zone was organized. The radioisotope concentration in sediments on the
bottom of individual parts of the Kiev reservoir adjacent to the accident
region was 10-7-10-8 Ci/kg and 10-10 Ci/L in water during the period
10-20 June 1986.
The radiation dose to which the aquatic organisms in the Kiev reservoir
were exposed will not have any serious effect on the population level.
Significant radiation effects on the aquatic ecosystem may be observed only in
the Chernobyl' plant cooling pond.
The hydrobionts populating the cooling pond were subjected to the
highest radiation burdens. For some species of water plant, the internal dose
received was as much as 10 rad/h, while near the bottom of the cooling pond
the average external exposure was 4 rad/h (at the end of May 1986).
According to expert evaluations, exposure levels of up to 10-2 rad/d
produce no noticeable effect on terrestrial ecosystems. Inside the 30 km zone
around the Chernobyl' plant, higher radiation levels were observed at
individual parts of the area contaminated by fallout: this may result in
significant changes in the state of radiosensitive plant species at these
points.
Radiation levels outside the 30 km zone cannot produce a noticeable
effect on the species of which the plant and animal associations are
composed.
The results obtained are preliminary in nature. Studies of the effects
of the Chernobyl' accident on living organisms and ecosystems are continuing.
7.4. Population exposures in the 30 km zone around the Chernobyl' plant
On the basis of an analysis of the radioactive contamination of the
environment in this zone, assessments were made of the actual and future
radiation doses received by the population of towns, villages, settlements and
other inhabited places. Following of these assessments, decisions were taken
to evacuate the population of Pripyat' and a number of other inhabited places:
135 000 people were evacuated.
As a result of these and other measures, it proved possible to keep
population exposures within the established limits.
The radiological effects on the population in the next few decades were
evaluated. The effects will be insignificant against the natural background
of cancerous and genetic diseases.
7.5. Data on the exppsure of plant and emergency service personnel. Medical
treatment
As a result of their participation in measures to combat the accident
in the first few hours after its occurrence, a number of plant and emergency
38
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service personnel received high radiation doses (more than 100 rem) and also
suffered burns during their efforts to extinguish the fire. All those
affected were given immediate medical attention. By 6:00 hours on
26 April 1986, 108 people had been hospitalized and in the course of the day a
further 24 persons out of those examined were admitted to hospital. One
person died from severe burns at 6:00 hours on 26 April and one of those
working at the damaged unit was not found. It is possible that he was working
in the area where structures had collapsed and there was high activity.
As a result of the early diagnosis procedures used in the Soviet Union,
within 36 hours persons in whom the development of acute radiation syndrome
was diagnosed as extremely likely had been identified for immediate
hospitalization. The hospitals selected were the clinical institutes in Kiev
closest to the site of the accident and a specialized unit in Moscow, the aim
being to provide the maximum amount of assistance and expert analysis of the
results of examinations
One hundred and twenty-nine patients were sent to Moscow in the first
two days. Of these, in the first three days 84 were identified as suffering
from degrees II-IV of acute radiation syndrome and 27 as degree I of acute
radiation syndrome. In Kiev there were 17 patients suffering from degrees
II-IV and 55 from degree I.
Details of the methods and results of the treatment of these patients
is given in the Annex.
The total number of fatalities caused by burns and acute radiation
syndrome among personnel stood at 28 at the beginning of July. None of the
population received high doses which would have resulted in acute radiation
syndrome.
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Table I. Daily release, q, of radioactive substances to the atmosphere
from the damaged unit (excluding radioactive inert gases)*.
Date
Days after'
the accident
q, Mc
26.04
0
12
27.04
1
4,0
28.04
2
3,4
29.04
3
2,6
30.04
4
2,0
01.05
5
2,0
02.05
6
4,0
03.05
7
5,0
04.05
8
7,0
05.05
9
8,0
06.05
10
0,1
09.05
14
~0,01
23.05
28
20.10-6
*Release evaluation error +5070.
It is composed of the error
of
the
dosimetric instruments, of
the radiometric measurements
of
the
radioisotopic composition of air and soil samples and.
of
the
error due to averaging the
fallout over the area.
**The values of q were calculated op ,6 May 1986 taking into
account radioactive decay. (At the time of the release on
26 April 1986, the activity was 20-22 MCi.) For the
composition of the release, see Table 2.
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Table 2. Assessment of the radioisotopic composition of the release
from the damaged unit*..
Isotope' -
Activity of release, MCi
26.04.86 I 06.05.86' ? ? )
Amount of activity.
released from the
reactor by 06:05.86, %
133Xe
5
45
Possibly up to 100
"InKr
0.15
" K
0,9
'III
4,5
7.3
20
112 i t.
4
1,3
15
1.14(.,
0,15
0,5
10
0,3
1,0
13
0,45
3,0
2,3
9` Zr
0,45
3,8
3,2
Ku
0,6
3,2
2,9
0,2
1,6
2,9
14ttn
0,5
4,3
5,6
141 ye
0,4
2.8
2,3
44 ce
0,45
2.4
2,8
49 Sr
0,25
2,2
4,0
Sr
0,015
0,22
4,0
0,1.10-3
0,13.10-3
3,0
239141
0,1.10-3
0,7.10-3
3,0
2401,t,
0,2.10-3
1.10-3
3,0
241
Pu
0,02
0,14
3,0
24 7 NI
0,3.10-13
2.10-6
3,0
242cm
0,3.10-2
2,1.10-2
3.0
2 PI Np
2,7
1.2
3,2
*Evaluation error +50%. For explanation, see footnote;
on Table 1.
Data on the activity of the main radioisotopes Measured
in the radiometric analysis.
Total release by 6 May 1986.-
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8. RECOMMENDATIONS FOR IMPROVING NUCLEAR POWER SAFETY
8.1. Scientific and technical aspects
In 1985 the Consultative Council for Co-ordination of Scientific
Research on Nucleai. Safety approved a "List of Priority Tasks", which
constitutes the basis for planning experimental and theoretical studies on
nuclear ?safety in the USSR aimed at providing a more detailed justification
for safety specifications, evaluating the actual level of nuclear safety and
enabling nuclear power plants commissioned before 1985 to be brought up to
that level in accordance with the specifications laid down.
After the accident at the Chernobyl' nuclear power plant, the status of
theoretical and experimental research on nuclear safety has been reviewed and
evaluated, and measures for extending, improving and intensifying it have been
developed.
Computer programs for analysing the safe behaviour of nuclear power
plants in all possible transient and accident regimes - including conditions
not anticipated at the design stage - are being improved, and modelling
systems and complexes are being developed.
Research on the possibility of building reactors with passive safety
systems -? so called "intrinsically safe" reactors, the cores of which cannot
be destroyed in any type of accident - is being expanded.
There will be an expansion of research on quantitative probabilistic
analysis of safety, on the analysis of risks from nuclear power and on the
development of a conceptual and methodological basis for optimizing radiation
safety and for comparing radiation hazards with other industrial hazards.
' 8.2. guanizational and technical measures
The system of monitoring and technical standards in force in the USSR
covers all the basic questions of nuclear safety and is continually
improving. In 1985 a Summary List and Development Plan for USSR nuclear power
regulations and standards was' drawn up under the auspices of the State Nuclear
Power Supervisory Board. (Cosatomehnergonadzor); this co-ordinates and directs
the activities of all official bodies involved in the development and
co-ordination of the corresponding scientific and technical documents.
A comparison between the existing Soviet document relating to nuclear
power station design and operation with similar foreign documents does not
reveal any major differences. In general, the nuclear safety standards in
force do not require revision. However, more careful verification of their
implementation in practice is necessary. The quality of training and
retraining of staff needs to be improved and design and construction staff
must verify more carefully the quality of plant components during manufacture,
assembly and adjustment during commissioning; their responsibility for the
subsequent effectiveness and safety of operating nuclear power plants must be
increased:
, 42
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Since the Chernobyl' accident, organizational measures have been taken
to improve the safety of nuclear power plants. These can be divided into two
stages.
The first stage, which was carried out before a detailed scientific and
technical analysis of the course of the accident had been made and in the
light of preliminary information from the site, relates to operating nuclear
power plants with reactors of the RBMIC type and involves operational measures
at those plants. The main purpose of these measures is to prevent any
recurrence of operating conditions such as those which immediately preceding
the accident.
The second stage relates to measures arising from the scientific and
technical analysis of the accident and includes steps aimed at improving the
safety of nuclear power plants of all types.
The measures that are planned should be adequate to ensure the safe
operation of nuclear power plants with reactors of the RBMK type.
? For power plants with other types of reactor, the intention is to carry
out safety enhancement measures foreseen earlier, which relate mainly to the
latest advances in science and technology, operating experience, the
possibility of diagnosing the condition of metal in piping and other plant
components, automatic process control systems, and so on.
With a view to raising the level of leadership and responsibility for
the development of nuclear power and to improving the operation of nuclear
power plants, an All-Union Ministry of Nuclear Power has been established.
A whole range of measures to improve State monitoring of nuclear safety
Is also to be carried out.
8.3. International measures
In the light of the Chernobyl' accident, the Soviet Union, paying due
regard to the international nuclear safety work currently being done and
desiring to strengthen international security further, has put forward some
proposals about the establishment of an international regime of safe nuclear
power development and the expansion of international co-operation in this
sphere. These proposals are contained in statements by the General Secretary
of the Central Committee of the Communist Party of the Soviet Union,
M.S. Gorbachev, on 14 May and 9 June 1986.
An international regime of safe nuclear power development would take
the form of a system of international legal instruments, of international
organizations and structures and also of organizational measures and
activities to preserve the health of the public and protect the environment in
the context of world-wide nuclear power activities. The establishment of such
a regime could be achieved by drawing up international agreements, signing the
corresponding international conventions and supplementary agreements, carrying
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out joint co-ordinated research programmes on nuclear safety problems,
exchanging scientific and technical information, selling up international data
banks and banks of material resources needed for safety purposes, and so on.
Funds could be set up, with the direct participation of international
organizations, for providing emergency assistance, including that required
with the urgent provision of the necessary special medical supplies and
dosimetric and diagnostic equipment and instruments and with the supply of
food, fodder and other material assistance. It is also necessary to set up a
system of early notification and provision of information in the event of
accidents at nuclear power plants - in particular, those with transboundary
consequences. Attention needs to be paid, moreover, to the question of the
material, moral and psychological damage associated with such accidents.
There is yet another aspect of nuclear safety, that of the prevention
of nuclear terrorism. A task of overriding importance in this connection is
the development of a reliable system of measures to prevent nuclear terrorism
In any form.
An important role in the establishment of an international regime of
safe nuclear power development must be played by the IAEA.
It is gratifying to note that initial steps have already been taken to
carry out the proposals in respect of the establishment of an international
regime of safe nuclear power development. Intensive work has begun on
preparations for the conclusion of two international conventions, relating to
early notification of nuclear accidents and to assistance in the event of
nuclear accidents and radiological emergencies. Certain aspects of the
expansion of international co-operation, in particular, the IAEA's research
Programmes on nuclear safety, are being actively discussed.
The proposals for establishment of an international regime of safe
nuclear power development are inextricably linked with problems of military
d?nte and nuclear disarmament. The Chernobyl' accident has demonstrated
once again the danger of nuclear energy getting out of control and has made
people aware of the devastating consequences which would ensue from its
military application or from damage to peaceful nuclear installations in the
course of military action. It is absurd, at the same time as discussing and
solving problems of the safe utilization of nuclear energy, to develop ways
and means of applying it in the most dangerous and inhuman way possible.
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9. THE DEVELOPMENT OF NUCLEAR POWER IN THE USSR
Owing to the extremely rapid development of nuclear power, a reduction
in the consumption of organic fuel by thermal power plants in the European
part of the country is planned in the Soviet Union's energy programme. The
contribution of oil to electric power production is to be cut by more than
half. Nuclear power should then cover most of the economy's increased
electricity requirements. There are plans for the maximum possible use of
nuclear fuel for centralized heating and industrial heat supply, and for the
creation of nuclear industrial complexes.
The Soviet Union is a pioneer in the peaceful use!: of atomic energy.
The world's first nuclear power plant, with a uranium-graphite channel-type
reactor, has been operating for thirty-two years. The subsequent programme
for the establishment in the USSR of so called demonstration power reactors
for nuclear power plants with relatively low power capacities made it possible
to select the most promising of these for further development and improvement.
The three types of reactor adopted in the USSR for the needs of the
country's growing nuclear power programme allow great flexibility and
reliability of energy supply, more efficient use of nuclear fuel resources
than would otherwise be possible, and are zilso well adapted to the special
requirements of a developing power engineering infrastructure.
The nuclear power plants being built in the USSR are based on the WWEH,
RHMK and fast breeder reactors. The first two are thermal reactors with
light-water coolant. The fast breeder reactors use liquid sodium as coolant
and are being built at present -with a view to full-scale industrial testing of
the technical solutions which have been adopted and the gradual future
development of a closed fuel cycle based on plutonium.
At present, nuclear power plants wiLh WWER .and RHMK reactors provide
the basic nuclear power production in the USSR. The country's installed
capacity has reached nearly 30 million kilowatts.. The Soviet nuclear power
plants are characterized by high operational availability. The utilization
factor of the installed capacity at nuclear power plants during recent years
has been relatively high.
In accordance with the "Main Lines of Economic and Social Development
of the USSR for 1986-1990 and up until the year 2000", it is expected that
nuclear power will be developed extremely rapidly in The European part of the
country and in the Urals. In 1985, power generation at nuclear plants reached
nearly 170 000 million kWh and by the year 2000 it will increase by 5-7 times.
This development means that it will be primarily nuclear plants that
provide the additional capacity needed for the energy systems of Lie European
part, relieving us of the need to build new thermal plants burning organic
fuel for base load operation.
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,The Soviet Union is also developing nuclear sources of heat supply
based_ on high-temperature gas cooled reactors. The construction of safe
plants with such reactors will make it possible to produce high-temperature
heat for a number of industrial technological processes.
The Soviet Union is actively participating in international
collaboration in the nuclear power field, co-operating effectively in the
competent bodies and commissions of the United Nations, in the IAEA, the World
Energy Conference and others.
The development of nuclear energy in the USSR is being carried out in
close co-operation with CMEA countries.
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.
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List of principal installations of the main block of the plant
No.
Installation or item
Measurement
unit
Unit
weight
in tOns
No. per
NPP unit
Reactor section
i
Graphite stack
Assembly
1850
1
2 -
System "S" metal structures
--"
12,6
1
3
System "OR" metal structures
"
280
1
4
System "E" metal structures
.
450
1
5
System "KZh" metal structures
"
79
1
6
System "A" metal structures
.
592
1
7
System "D" metal structures
236
1
8
Drum-type steam separator
In
.278
4
9
Main circulation pump TsVN-8
.
67
8
10
Main circulation pump electric motor
19
33
11
Main isolating gate valve (diameter 800)
91
5.7
8
12
'Intake header
.
41
2
13
Pressure header
99
46.0
2
14
Distributing group header
.
1.3
44
15
Lower water communication lines
Set
400
1
16
Steam-water communication lines
.
450
1
17
Downcomers (diameter 300)
.
16
1
17a
Primary coolant circuit pipes (diameter 800)
19
350
1
18
Refuelling machine
.
450
1
19
Overhead crane of central hall Q50/10ts
pc
121
1
20
Overhead crane of main circulation pump room
Q50/10ts
.
176
2
21
Supply fan, type VDN at level + 43.0
.
3.5
30
22
Exhaust ventilator at level + 35.0
19
3.5
50
23
Controlled leakage tank
.
1.4
2
24
Controlled leakage heat exchanger
.
0.2
2
25
Scheduled preventive maintenance
19
25
4
26
,tank
Metal structures and pipes of the accident
confinement zone
Set
270
1
27
Check valves of the lower water communication
line room
19
2.5
11
28
Accident confinement system release valve
pc
2
8
29
Accident confinement system condensers
3.7
36
30
Carriage container
.
146
1
31
Crane in gas activity reduction system room
Q30/Sts
45
1
Carbon steel pipes
Set
1170
1
Stainless steel pipes
.
760
1
Machine room
32
Turbo-unit K-500-65/3000
PC
3500
2
33
Moisture separator/reheater SPP-500
19
15
8
34
Low-pressure heater
11
37.5
4
35
First stage condensate pump units
.2.5
6
36
Overhead crane of machine room Q 125 ts
11
211
Carbon steel pipes
Set
3825
Stainless steel pipes
1300
1
37
De-aerator
PC
4.5
2
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AIM) HS11 110.1
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?P's
(X) 4
gurtria.A. A 1.11. AL A6
*?
_
A
Oft
II
ISMS
74044
?
I t Iwt I -
. I
- I
0.444.4 Z:Tif
wn
?r?r
000 40041
?
(A) (I) (A)
4) 60 Qi
1
73450
4) () od
09 CO Q0
3i
0 ft)
Fig. 1? Cross?section through the main structures of a power plant with an RBMK-1000 reactor (with localization zone)
Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
? SIB -.Ca:C=0
41000
6
Secondary
cooling water '
v.
c.
400
13
12
Fig. 2. Schematic diagram of the emergency core cooling system (ECCS):
1. Reactor; 2. Steam separator; 3. Suction header; 4. Main circulation pump; 5. High pressure header; 6, Pressure suppression pond;
7. ECCS vessels; 8. ECCS pumps for cooling the damaged half of the reactor; 9. Heat exchanger; 10. Clean condensate tank;
11. FCCS pumps for cooling the undamaged half of the reactor; 12. Deaerator; 13. Feed pump.
KIN() 1Sfl 1VLJIJJO 110,1
0-1-0009Z001-0006600-60c1C1I-V10 91-/L0/?1,0Z eseale Jod panaddv pue pewssepea
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?
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0
,
1 1
,
1
o
0
0.
,
_
o
,
o
o
a
o
.------i--
?
r
8
I
:6,
0.,
_
1
I
51
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0
-z
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5
t, days
1()
Fig. 5. Variation of activity and temperature of the fuel with time
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Unit II
Unit III
Unit IV
Metal dividing wall
Concrete dividing wall -
Engineer's wall
Burial structures
Zone of destruction of
principal structures
Principal structures
Fig. 6. Diagram showing one scheme for theisolation and
encasement of unit 4 (horizontal cross?section)
71,30
4--
51,00 113,01_ 30,00 112.00112,001 24,00 17112,001 30,00
Fig. 7. Diagram showing one scheme for the isolation and
encasement of unit 4 (vertical cross?section)
55
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Fig. 8. General view of one scheme for the isolation and encasement of unit 4
?
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PART II., ANNEXES
ANNEX1
WATER-GRAPHITE CHANNEL REACTORS AND OPERATING
EXPERIENCE WITH RBMK REACTORS
1. Water-graphite channel reactors and operating experience with RBMK
reactors.
1.1. Water-graphite channel reactors use ordinary water as a coolant and
graphite as a moderator. The distinctive features of channel reactors are:
the absence of a pressure vessel, the relative simplicity of the design, the
extensive scope for channel-by-channel inspection and control, the possibility
of refuelling while the reactor is operating, the flexibility of the fuel
cycle, and the practically unlimited potential for increasing the power by
means of standard structural elements.
The first power reactOr in the USSR was a channel-type reactor - the
water-graphite reactor of the First Atomic Power Station, which had an
electrical capacity of 5 NW and was started up in June 1954 at Obninsk, near
Moscow.
1
. _
The 'experience acCumulated:constitc?on and operation of the
. .
First Atomic Power Station Was utilized in planning the Beloyarsk Nuclear
,
Power Plant'(NPP) (1964,-300 MW)..
The further development of the water-graphite reactor concept in the
USSR led to the construction of the high power channel-type boiling-water
reactor RBMK-1000 with an electrical capacity of 1000 MW, which, along side
the WWER-1000 reactor, became the basic reactor for large-scale nuclear power
production in the USSR.
The commissioning of the, first RBMKT1000 reactor At the Leningrad NPP
in 1973 marked the inception of a series of reactors of this type.
The broad programme of construction of RBMK-1000 reactors carried out
in the 1970s led to the commissioning, over the period 1973 to 1985, of
14 reactors (4 reactors each at the Leningrad, Kursk, and Chernobyl' NPPs and
2 at the Smolensk NPP) with a total installed electrical capacity of 14 Cu.
In each new generation of reactors, improvements aimed at increasing their
reliability and safety were introduced.
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The knowledge and experience gained from operating generating units
with RBMK-1000 reactors revealed their intrinsic safety margins and made it
possible to design, on the basis of this type of reactor, the even more
powerful RBMK-1500 reactor, with an electrical capacity of 1500 MW, which was
brought into service in 1983 at the Ignalinsk NPP and by early 1985 reached
its design output.
Channel-type reactors have advantages which facilitate the solution of
safety problems. These advantages include:
- their convenience for carrying out individual checks on the state
of the fuel elements and fuel assemblies and on channel integrity;
- the possibility of operationally exchanging failed fuel assemblies
without shutting down the reactor;
- the reduced hazard from primary circuit pipe breaks due to the
increased number of coolant loops and the corresponding reduction
in tube diameters;
the design option of increasing the unit power of the reactor
without making the emergency core cooling system more complicated.
On the other hand, certain specific features of channel-type graphite
reactors cooled with boiling water call for fundamentally new solutions to be
found in developing the safety systems. These features include, in particular:
- the large steam volume in the coolant circuit, which considerably
slows the rate of coolant pressure reduction after an accidental
pipe break;
the potential occurrence of a positive void reactivity effect,
which to a large extent governs the behaviour of the
neutron-flux-determined power of the reactor in emergencies due to
interruptions of the coolant circulation through the core;
the large amount of thermal energy accumulated in the metal
structures and graphite stack of the reactor, which affects the
decline in thermal power after a response of the safety system
(scram).
1.2. As of now, the cumulative service life of RBMK reactors within the
nuclear power system is approaching a total of 100 reactor-years.
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On the basis of analysis and generalization of this operating
experience, the various reactor components and plant systems and operating
regimes are continually being updated and improved. As a result, a large
number of measures have been developed and introduced with a view to
increasing the reliability and operating safety of nuclear power plants. The
most important of these measures are the following:
- modernization of the design of the isolating and regulating valves,
the ball detectors of the flowmeters and the fuel channel shut-off
plugs;
- optimization of the routing of the steam/water piping and the steam
discharge tubes of the drum separators;
- improvement of the drum separator internals;
- improvement of the main circulation pumps and their auxiliary
systems;
introduction of prognostic programmes for operational calculations,
of programmes for recording the state of equipment during
emergencies and of diagnostic programmes for monitoring the process
systems;
- development and introduction of local automatic control systems and
local emergency protection systems based on in-core sensors;
justification and experimental operation at one of the generating
units of fuel assemblies with an initial enrichment of 2.4%;
- development of afterheat removal systems allowing extended
continuous repairs of reactor equipment and components.
These measures and others carried out at the generating units have
ensured the reliable and safe operation of nuclear power plants with RBMK
reactors, some statistics on which are given in Table 1.2 showing the
production of electrical energy by nuclear power plants over the operating
period 1981-1985 and the installed capacity factor for 1985.
The maximum values obtained for the installed capacity factor in 1985
were 91% at unit 4 of the Leningrad NPP, 90% at unit 2 of the Chernobyl' NPP
and 87% at unit 1 of the Leningrad NPP.
Generalization from the operating experience accumulated and from the
scientific research and development work carried out has indicated some ways
of increasing the effectiveness of RBMK generating units, including:
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increasing ,the power_ of existing generating units, improving and optimizing
the operating regimes , of the units, introducing automatic reactor and
equipment protection systems, improving the conditions under which repairs are
carried out on the reactor plant and increasing the maintainability of its
individual components.
As, is well 'known, the limiting parameters for the power of RBMK
reactors are the fuel temperature, the temperature of the graphite stack and
metal structures, and the burnout ratio of the fuel channels. At existing
reactors, these parameters are lower than the maximum permissible. For
example, at the nominal thermal power of the reactor the maximum power of a
fuel channel is about 2600 kW, where 3000 kW would be permissible, and the
maximum temperature of the graphite stack is 923 K (6500C), where 1023 K
(7500C) would be acceptable; the maximum temperature of the metal structures
is 573 K (3000C), where 623 K (3500C) would be allowed, and the burnout
ratio is not less than 1.35. The main items of equipment in the turbine room
of RBMK generating units (turbogenerators, unit transformers, de-aerators,
condensate and feed pumps) also have a safety margin of ?10% with respect to
power.
These demonstrated safety margins have made it possible to justify
operating the generating units at a higher power level, while the full-scale
tests carried out in several generating units at powers of up to 107% of
nominal have confirmed this possibility.
Table 1.1. The main technical characteristics of nuclear power plants (NPPs)
with RBMK-type reactors are given in Table 1.1.
Characteristics
RBMK-1000 RBMK-1500
Electrical power (MW) 1000 1500
Thermal power (MW) 3200 4800
Steam output (t/h) 5800 8800
Steam parameters before turbines:
Pressure (kgf/cm2) 65 65
Temperature (?C) 280 280
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Table 1.2. Performance data for the operation of NPPs with RBMK-1000 reactors.
Performance data Leningrad Kursk Chernobyl' Smolansk
NPP NPP NPP NPP
Installed capacity as of 1/1/86 (MW) 4000 4000 4000 2000
Electricity production for the
period 1981-1985 (109 kW.h)
140.4
82.4
106.6-
23.4
Installed capacity factor for
1985 (%)
84
79
83
76
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ANNEX 2
2. DESIGN OF THE REACTOR PLANT
The reactor plant is designed to produce dry saturated steam at a
pressure of 70 kgf/cm2 (= 7 MPa). It consists of the reactor proper with
its monitoring, control and protection systems, and the piping and equipment
of the multiple forced circulation loop (primary coolant circuit).
2.1. Reactor
The RMBK power reactor is a heterogeneous thermal neutron channel-type
(pressure tube) reactor, in which graphite is used as the moderator, while the
coolant is light water and a steam-water mixture circulating through vertical
channels passing through the core.
The reactor core (1) takes the shape of a vertical cylinder with an
equivalent diameter of 11.8 m and height of 7 in (see Fig. 2.1). It is
surrounded by lateral and end graphite reflectors 1 and 0.5 in thick,
respectively; The core is composed of fuel channels with the fuel assemblies
inside them, a graphite moderator, channels with neutron absorber rods
(control rods) and the sensors of the monitoring system. Some of the channels
in the tore are made of a zirconium alloy. The graphite stack consists of
blocks assembled into columns with axial cylindrical openings into which the
fuel channels are inserted. The fuel channels are located in 1661[*] cells in
a square lattice with a 250 mm pitch. The channels of the control and
protection system (CPS) number 211 and are arranged in the same way as the
fuel channels in the central openings of the graphite stack columns (the
arrangement of the channels is shown in Fig. 2.1a).
The grapite stack is located in a leaktight cavity (reactor space)
formed by the cylindrical cowling (2) and the plates of the upper (4) and
lower (3) metal structures. To prevent oxidation of the graphite and to
improve heat transfer from the graphite to the fuel channels the reactor space
Is filled with a helium-nitrogen mixture with a volumetric composition of
85-90% He and 15-10% N2. To prevent the possibility of helium leaking from
the reactor space the inside cavities of the metal structures and the space
around the cowling are filled with nitrogen at a pressure 50-100 mm H20
(~ 0.5-1.0 kPa) greater than the pressure in the reactor space.
The reactors of the first construction stages of the Leningrad, Kursk
and Chernobyl' nuclear power stations contain 1693 fuel assemblies and
179 CPS channels.
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The fuel channels are housed in tube ducts welded to the metal
structures (5). The upper and lower metal structures and the water-filled
annular tank (6) around the cowling serve as biological shielding for the
rooms surrounding the reactor. The coolant (water) is fed in from below to
each fuel channel through separate pipes. As it rises and flushes the fuel
elements, the water heats up and partially evaporates; the steam-water mixture
is led off from the top of the channels likewise through separate piping.
Nuclear fuel is reloaded without a reduction in reactor power by means
of the refuelling machine.
Under steady-state operating conditions the intensity of the refuelling
when the reactor is operating at nominal power is 1-2 assemblies per day.
The reactor is equipped with a control and protection system (CPS) and
with monitoring systems which transmit information on the state of the core
and the operation of various components, as well as sending the necessary
signals to the CPS and the emergency signalling system.
Main characteristics of the reactor
Coolant flow through the reactor, t/h
37.6 x 103
Steam pressure in the separator, kgf/cm2
70
Pressure in the group pressure headers, kgf/cm2
82.7
Mean steam content at the reactor outlet, %
14.5
Coolant temperature, 0C
Inlet temperature
270
Outlet temperature
284
Maximum channel power, with allowance for 10%
power distortion, kW
3000
Coolant flow rate in maximum power channel, t/h
28
Maximum steam content at channel outlet, %
20.1
Minimum critical power margin
1.25
Core height, mm
7000
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Core diameter, mm 11 800
_Fuel lattice pitch mm 250 x 250
Number of fuel channels 1661
2.1.1. Design of the fuel assembly and fuel element
The fuel assembly of the RMBK 1000 reactor consists of the following
main parts (Fig. 2.2):
Two fuel sub-assemblies (1);
- Supporting rod (2);
Guiding tail and nose pieces (3 and 4);
- Nuts (5).
The fuel assembly is 1015 mm long.
Each sub-assembly consists of 18 fuel elements, a casing and
18 pressure rings.
The fuel element (2.2a) consists of the cladding (6), fuel column (7),
holding spring (8), plug (9) and end piece (10).
The material of the cladding and end pieces is a zirconium alloy with
1% niobium (alloy. 110). The spring is made of Ts2M zirconium alloy.
The outer diameter of the cladding is 13.6 mm and the minimal thickness
0.825 mm.
As the fuel use is made of sintered uranium dioxide pellets. The
pellets are 11.5 mm in diameter and 15 mm high; to reduce the heat expansion
of the, fuel column the pellets are concave at the end. The mean mass of fuel
in a fuel element is 3600 g, the minimum density of the pellets is
10.4 g/cm3, and the diametric gap between the fuel and the cladding is
0.18-0.38 mm.
?The fuel elements are made leaktight by resistance butt welding of the
nose piece on to one end of the cladding tube and of plug on to the other.
The ? initial medium under the cladding is helium at a pressure of
? 1 kg/cm2 (0.1 MPa). The fuel column in the element is held in place by
the spring with a constrictive force of about 15 kg.
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The casing consists of a central tube 15 mm in diameter with a wall
thickness of 1.25 mm, an annular grid (11) and 10 spacer grids (12). The
central tube and end grid are made of a zirconium alloy with 2.51. niobium
(alloy-125), while the spacer grids are made of stainless steel.
By means of two flairings the central tube is joined to the end grid in
such a way that there is no possibility of an axial air gap at the join, and
twisting of the grid with respect to the tube is also prevented. To keep the
sub-assemblies in position and prevent them twisting with respect to one
another, the casing tubes are fitted with special grooves. The spacer grids
are fixed to the central tube at intervals of' 360mm. Each grid is secured by
insertion of the projecting end of the central sleeve into two grooves on the
tube in such a way that it can move along the tube if there is a small
azimuthal air gap.
The spacer grid is assembled from individual shaped cells (12 cells in
the peripheral row and 6 in the inside row), the central sleeve and an
encircling rim. The parts of the grid are joined together by resistance spot
welding. The openings for the fuel elements in the grid are 13.3 mm in
diameter. On the rim of the grid there are projections making it easier to
load the assembly into the channel. The diameter across the rim projections
Is ? 78.8 mm.
The cells are made of tubing with a wall thickness of 0.35 mm; the
central sleeve is made of tubing with a wall thickness of 0.5 mm, and the rim
from tubing with a 0.3 mm in wall thickness.
The fuel elements are secured to the end grid by means of pressure
rings made of stainless steel. The securing system cannot be taken apart,
since the pressure rings deform when the fuel elements are secured.
The design of both fuel sub-assemblies is identical.
When the fuel-assembly is being put together, the nose piece, the two
sub-assemblies, and the tail piece, which is fixed with a nut, are mounted on
the central rod. The nut is prevented from unscrewing by means of a pin.
Two types of fuel-assembly are inserted in the reactor: a working
assembly and an assembly for use as a monitor for the power density Aover the
core radius) which is different from the working assembly in terms of the
design of the tie rod. The latter is hollow and consists of a tube width a
12 mm outside diameter and wall thickness of 2.75 mm, and a plug, both made of
zirconium alloy (alloy-125), a steel-zirconium transition piece and an
extension tube made of stainless steel.
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2.1.2. Fuel channel (Fig. 2.3)
The fuel channel is intended to house the fuel assemblies with the
nuclear fuel and to control the flow of coolant. The casing of the channel is
a welded structure consisting of a middle and end part. The middle (2) is
made of zirconium alloy (Zr + 2.5% Nb) and composed of a tube 88 mm in outside
diameter with a wall thickness of 4 mm, an upper (1) and lower (5) end piece
made of corrosion-resistant tubing (steel 08 Crl8NilOTi). The middle part is
joined to the ends by means of special steel-zirconium transition pieces
(3, 4).
The transition joints - corrosion-resistant steel-zirconium alloy - are
manufactured by means of vacuum diffusion welding (Fig. 2.3(a)).
The transition joints are designed to produce programmed configurations
and stresses in the area of the joint that guarantee strength and reliability
under operating conditions. The inside part of the transition is made of
zirconium alloy, while the outside part around it is made of corrosion-
resistant steel. During the diffusion welding a thin layer of mutually
diffusing products forms on the contact surface of the parts being joined
together. The quality of the diffusion welding is checked by ultrasonic flaw
detection and metallographic devices. As part of the fuel channel the
transition pieces are also tested for helium leaktightness and hydraulic
pressure.
The channel tubes are joined to the zirconium parts of the transition
pieces by electron-beam welding. To improve the corrosion properties of the
welded joints they undergo additional strengthening and heat treatment.
The steel parts of the transition pieces are welded to the top and
bottom parts of the fuel channel by argon welding. A metallic coating of
aluminium is applied to the outer surfaces of the steel parts in the channel
to protect them against corrosion.
To improve heat flow from the graphite block to the channel, slotted
graphite rings 20 mm high are fitted onto the middle of it and positioned very
closely together along the channel so that every other ring is directly in
contact, by means of its lateral surface, either with the pipe (7) or with the
inside surface of the block (6), as well as being in contact at their ends.
The minimum gaps between the channel and ring - 1.3 mm - and ring and
block - 1.5 mm - are designed to prevent wedging of the channel in the stack
through radiation-induced thermal shrinkage when the reactor is in operation.
The channel body is -housed in the reactor in tube ducts (3, 4) welded
to the top and bottom metal structures (Fig. 2.4). It is attached immovably
to the upper duct by means of a thrust collar and filament seam made by argon
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arc welding (1). The lower part of the body is welded to the metal structure
duct, being joined to it through the bellows compensating unit (2); this makes
it possible to compensate for any difference in thermal expansion of the
channels and metal structures, as well as ensuring reliable leaktightness of
the reactor space. The channel body is designed to operate safely for
30 years, but whenever necessary a defective channel body can be taken out of
the reactor and replaced by a new one with the reactor shutdown.
The fuel assembly with its fuel elements (5) is mounted inside the
channel on a suspension (6), which keeps it in the core and enables the
refuelling machine to replace a spent fuel assembly without stopping the
reactor.
The suspension is fitted with a closing plug (7), which is mounted in
the housing of the upper duct. This plug hermetically seals the inside of the
duct by means of a ball-type shutter fitted with a sealing washer. The
unsealing operations during refuelling are carried out by, the, refuelling
machine using remote control.
2.1.3. Control channels (Fig. 2.4)
These channels are intended to contain the control system rods,
vertical power density monitors and ionization chambers. The middle of the
channel (3) is made of a zirconium alloy (Zr + 2.5% Nb) and constitutes a tube
88 mm in diameter and with a wall thickness of 3 mm. The upper (1) and
lower (4) end parts are made of corrosion-resistant piping (steel
08 Crl8NilOTi). The middle part is joined to the 'end tubes by means of
steel-zirconium transition pieces similar to those used for the fuel
channels. The channels are secured immovably to the upper tube duct by means
of a thrust collar and a filament seam, and to the lower duct via the bellows
compensating unit. The CPS channels in the upper part have heads (5) designed
for the attachment of actuators and for supplying cooling water to the
channel. Graphite sleeves (6) are placed over the channel and provide the
requisite temperature conditions for the graphite column. At the bottom of
the channel is a throttle device (2), which ensures that the 'channel is
completely filled with water.
Placing of the control channels in the graphite columns independently
of the fuel channels guarantees their preservation and, consequently, the
efficiency of the control elements contained in them in the event of possible
accidents due to rupture of the fuel channels.
2.1.4. Metal structures of the reactor (Fig. 2.1)
Thelateralbiologicalshieldingtank(f)takesthe,form of a
cylindrical reservoir with an annular section 19 at in outside diameter and
16.6 m in inside diameter: it is made of low-alloy steel sheeting of the
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pearlite class (10 CrSiNiCu) 30 mm thick. Inside the tank is divided into
16 vertical leaktight compartments filled with water, the heat from which is
removed by the cooling system. The top metal structure (4) is a cylinder 17 m
in diameter and 3 m high. The upper and lower plates of the cylinder are made
of steel (10 CrNilMo) 40 mm thick welded to the lateral shell by means of
leaktight welds, and welded to each other by means of vertical strengthening
fins. The holes in the top and bottom plates are for the welded-in tube
ducts (5) holding the fuel and control channels. The space between the tubes
is filled with serpentinite (a mineral containing bound water of
cystallization). The metal structures are mounted on 16 roller-type supports
attached to the projection of the annular part of the lateral biological
shielding and bear the weight of the loaded channels, the floor of the central
hall and the piping of the upper steam-water and water communication lines.
The bottom metal structure (3), which is 14.5 m in diameter and 2 m
high, is similar in design to the top structure. It is loaded by the graphite
stack mounted on top of it together with the supporting units and lower water
communications. The number and arrangement of the lower fuel and control
channel ducts welded to the top and bottom of the lower metal structure are
the same as in the uppe'r. structure. The cavity inside it is filled with
serpentinite. The supporting metal structure on which the lower metal
structure is mounted is composed of plates with reinforcing fins 5.3 m high
which intersect at the centre of the reactor and are perpendicular to each
other (7).
The cylindrical shroud (2) is a welded shell with an outside diameter
of 14.52 m and height of 9.75 in made of steel sheeting (10 CrNilMo) 16 mm
thick. To compensate for longitudinal heat expansion the shroud is fitted
with a lens-type compensator. The shroud, together with the top and bottom
metal structures, forms the closed reactor space.
The metal structure of the top covering (8) has an opening for the
insertion of the fuel and other special channels. It is covered over by a
removable floor (9) consisting of individual slabs. The floor acts as
biological shielding for the central hall and, furthermore, serves as heat
insulation for it. The floor consists of upper and lower slabs and blocks
resting on the fuel and reflector channel ducts. The slabs and blocks are
metal structures filled with iron-barium-serpentinite cement stone.
Air is extracted from the central hall through gaps in the floor and
then passes to the ventilation shafts. The air cools the floor and prevents
the possibility of radioactivity releases entering the hall from the room
containing the steam-water communications.
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2.1.5. Graphite stack (Fig. 2.1)
The graphite stack (1) is assembled on the lower metal structure inside
the reactors space. It takes the form of a vertical cylinder made up of
2488 columns of graphite blocks with a density of 1.65 g/cm3. 'The blocks
are shaped like parallelepipeds with a 250 x 250 mm section and height of
600 mm. The mass of the stack is 1700 t. There are openings 114 mm in
diameter along the axis of the blocks, forming ducts in the columns to hold
the fuel channels and CPS channels. Each graphite column is mounted on a
steel base plate (10), which in turn rests on a cup welded to the top plate of
the lower metal structure. The graphite stack is made secure against movement
in a radial direction by means of rods positioned in the peripheral columns of
the lateral reflector. At the bottom the rod is welded to the supporting cup,
while at the top it is joined immovably to the tube duct welded to the bottom
plate of the upper metal structure. The hollow rod, made of corrosion
resistant steel (08 Crl8NilOTi) piping, holds the channel for cooling the
reflector blocks. The heat released in the stack is removed basically to the
fuel channels and partially to the CPS channels. The presence of firm-contact
rings on the channels and the helium-nitrogen mixture with which the channel-
ring and ring-block gaps are filled keep the stack at a temperature not
exceeding 7000C.
In the case of the graphite blocks the highest temperature zones are to
be found on the block edges, while the lowest temperatures are found on the
inner surface of the vertical openings into which the fuel and other channels
are placed. The highest temperature is found in by the blocks located in the
middle of the centre part of the core.
The greatest temperature differential - between the edge and inner
surface of the opening - is to be found in the block with the fuel channel and
amounts to - 1500C.
2.1.6. Biological shielding
The biological shielding of the fourth unit reactor of the Chernobyl'
nuclear power station has been designed in accordance with the requirements in
force in the USSR - "Radiation Safety Standards NRB-76" and "Health
Regulations for Designing and Operating Nuclear Power Plants SP-AEhS-79".
The dose rate for external exposure in the central hall and serviced
buildings adjoining the reactor vault do not exceed 2.8 x 10-2mSv/h
(2.8 mrem/h). During refuelling, at the time when the spent fuel assembly is
removed and passed through the floor of the central hall, the gamma dose rate
close to the refuelling machine briefly rises to 0.72 mSv/h. In the room
containing the water communication lines below the reactor the shielding
ensures that the neutron flux density drops to values at which there will_ be
no appreciable activation of the piping and structures. It is only permitted
to enter that room when the reactor has been shut down.
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Shielding against radiation from the coolant in the piping and
equipment of the main circuit makes it possible to carry out repair and
adjustment operations while the reactor is in operation; for example,
channel-by-channel adjustment of the coolant flow by means of multipurpose
valves fitted in the group headers, repairs to the electric motors of the main
circulation pumps, and so on. Radiation heat release is reduced to values at
which the temperature of the supporting metal structures (top, bottom and
tank) and the reactor shroud is not more than 3000C, which makes it possible
to use low-alloy steel.
The fast neutron fluence with an energy of more than 0.1 MeV reaching
the reactor shroud and sheeting of the metal structures close to the core has
not exceeded 1020n/cm2 in 30 years of operation.
The shielding designed takes the following form (Fig. 2.1).
Mounted on each graphite column, between the end reflectors 500 aim
thick and the upper and lower metal structures, are steel blocks (10) (the
lower ones are 200 mm thick and the upper ones 250 mm) designed to reduce the
fast neutron fluence onto the metal structures supporting the load, as well as
to reduce the energy released in them.
The space between the tubes in the top and bottom metal structures is
filled with serpentinite (3, 4), which makes it possible to reduce the length
of the fuel channels and the overall dimensions of the building.
Above the steam-water communication lines is a protective covering
(floor of the reactor hall), the central part of which - the slab flooring
(9) - is made up of a set of blocks resting on the tops of the channel ducts.
These blocks are made of iron-barium-serpentinite cement stone. The overall
thickness of the covering is 890 mm. The upper flooring protects the central
hall against radiation from the reactor and from the piping containing the
radioactive coolant, and together with the refuelling machine container
reduces the intensity of the radiation when unloading spent fuel assemblies.
The peripheral part of the upper covering (8) constitutes metal cases 700 mm
high filled with a mixture of pig iron shot (86% mass) and serpentinite.
In a radial direction the lateral reflector consists of four graphite
blocks, with a mean thickness of 880 mm. The annular water tank (6) lying
behind the reactor shroud reduces the radiation fluxes to the walls of the
reactor vault (11), which are made of building concrete (density 2.2 t/m3
and wall thickness 2000 mm). The space between the tank and the walls is
filled with ordinary sand (12).
The thicknesses and composition of the materials of which the RBMK
reactor shielding is made in the main directions away from the core are shown
in Table 2.1.
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Table 2.1. Thickness of shielding materials (in a direction away from the core
centre) (mm).
Material
Direction
Upward
downward
radial
Graphite (reflector) (mm)
500
500
880
Steel (protective plates and sheeting
of the metal structure) (mm)
290
240
45
Serpentinite filling (1.7 t/m3)(mm)
2800
1800
-
Water (annular tank)(mm)
-
-
1140
Steel (metal structures)(mm)
40
40
30
Sand (1.3 t/m3)(mm)
-
1130
Heavy concrete (4.0 t/m3)(mm)
890
-
-
Building concrete (2.2 t/m3)(mm)
-
-
2000
A reduction in the intensity of radiation streaming through the
gas-filled channels (for temperature sensors, neutron flux detectors and
ionization chambers) or channels with less effective shielding (steam-water
mixture in fuel channels) is attained by inserting shielding plugs made of
steel or graphite (Fig. 2.5). The annular gaps between the channels and the
guide tubes are closed by means of shielding sleeves (Fig. 2.6).
The gas piping which passes through the shielding structures is made
with bends (No. 13 in Fig. 2.1).
To prevent neutron streaming and gamma radiation, as well as to reduce
the activation of the structures in the area below the reactor, the displacers
in the CPS channels are filled with graphite (Figs 2.17 and 2.30).
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za..33333
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2
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Fig. 2.2
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Zse?
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(Thio
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REHEC fuel channel
'Mates feed
Fig. 2.3
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77
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C\2
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Fig. 2.S. Positionof shielding plug in fuel chmmtel: (1) steel sleeves;
(2) helical steel plug; (3) channel tube; (4) serpentinite filling
Fig. 2.6. Arrangement of shielding sleeves in upper reflector area:
(1) graphite sleeves; (2) steel shielding block;
(3) graphite reflector
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2.2. Primary circuit (multiple forced circulation circuit)
(Fig. 2.6) [?]
The purpose of the primary circuit is to supply water to the process
channels and to remove the steam-water mixture, which forms in them as a
result of the heat taken up from the fuel assemblies, for subsequent
separation of the steam. It consists of two loops, similar in their
arrangement and equipment, which function in parallel and indendently; each
removes heat from half of the reactor's fuel assemblies. A loop includes:
2 drum-type steam separators (int. diam. 1. 2600 mm), downpipes (325 x 16),
4 main circulation pumps, main circulation pump suction pipes
(int. diam. = 752 mm) and fittings, main circulating pump pressure header
(int. diam. = 900 mm), distributing headers (325 x 15 mm) with isolating and
regulating valves, water lines (57 x 3.5 mm), process channels and steam lines
(76 x 4 mm). (A diagram of the primary circuit fittings is shown in
Fig. 2.7). [Missing from original.]
Water from the suction header (1) passes through four pipes to the
main circulating pumps (2). Under normal operating conditions at normal power
three of the four main circulating pumps are in operation, with one held in
reserve. Water leaves the main circulating pumps at a temperature of 2700C
at a pressure 82.7 kgf/cm2 through pressure pipes, in each of which are
installed in sequence a non-return valve, a gate valve and a throttle valve,
and then flows into the main circulating pump pressure header (3), from where
it passes through 22 lines into the distributing headers (4), which have
non-return valves at their inlets, and then along individual water lines (5)
into the process channel inlets (6). The flow rate through each process
channel is determined by means of isolating and regulating valves in
accordance with the flowmeter readings. As it passes through the process
channels, the water surrounding the fuel elements is heated to saturation
temperature, partially evaporates (14.5% on average) and the steam-water
mixture at a temperature of 284.5?C and a pressure of 70 kgf/cm2
(-7 MPa) flows through the individual steam lines (7) into the separators
(8), where it is separated into steam and water. In order to keep the levels
the same, the' separators are interconnected with separate shunts for water and
steam., Saturated steam passes through the steam collectors to the turbines.
The water which has been separated out is mixed at the separator outlets with
feed water, and flows through 12 downpipes (from each separator) into the
suction header at a temperature of 2700C; this provides the cavitation
margin required by the main circulating pumps.
The temperature of the water flowing into the suction header depends
on the rate of steam production of the reactor unit. When this decreases, the
temperature increases somewhat-because of the changing ratio of water from the
drum separators, at a temperature of 2840C, and feed water, at a temperature
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of 165?C. When the reactor is being powered down, the flow rate through the
primary circuit is controlled using throttle-type control valves so that the
temperature at the main circulation pump inlet maintains the necessary
cavitation margin.
2.3. Special control channel cooling circuit
There is a special, iadependent cooling circuit for the side screen and
the control channels, vertical power density monitors and the startup
ionization chambers. The water circulates under gravity, i.e., because of the
difference in level between the upper (storage) and rower (circulation)
tanks. Cooling water at 40?C flows from the upper tank through a header
along individual lines to the channel end-plugs, and continues downwards
removing heat and warming up in turn to a temperature of 65?C. It then
passes through a discharge header into heat exchangers, where it is cooled to
40?C, and collects in the lower tank, from which it is pumped back up into
the upper tank. The mean flow rate through the control channels is 4 m3/h
and the overpressure at the channel end-plugs is 3.5 kgf/cm2. The flow rate
through each channel is controlled using isolating and regulating valves in
accordance with the flow meter readings.
2.4. The gas circuit
Under normal operating conditions, a helium-nitrogen mixture flows at
200-400 nm3/h [sic] at an overpressure on entering the reactor space of
50-200 mm head of water equivalent (0.5-2.0 kPa) through pipes which pass
through the lower part of the metal structure, it is removed through the
process channel failure monitoring system pipes and through special channels
which remove the gas from the piping sectors of the upper part of the metal
structure. The gas mixture then passes through a condenser, a three-stage
scrubbing system, its flow rate is controlled by throttle and it returns to
the reactor space. The gas is circulated by means of compressors.
The gas scrubbing system consists of a set of contact catalysers,
scrubbing and dewatering units and cryogenic cooling system units. In the
contact catalyser, hydrogenation with H2 takes place at a temperature of
-1600C, with the formation of water vapour and combustion of CO to CO2
and the release of heat. The reaction takes place in an oxygen atmosphere in
the presence of a platinum catalyser. After passing through the contact
catalyser, the gas passes through a refrigerator and dehumidifier and then on
into the scrubbing and dewatering unit, which is equipped with zeolite and
mechanical filters. Adsorption takes place and CO2, H3, C2 and water
vapour impurities are scrubbed from the helium-nitrogen gas, which then passes
to the cryogenic cooling unit. Any impurities remaining in the gas are
removed in this unit by dephlegmation at a temperature of -1850C.
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2.5. Basic physics data
The RBMK-l000 nuclear power reactor is a heterogeneous channel-type
thermal-neutron reactor which uses uranium dioxide of low enrichment in 235U
as fuel, graphite as moderator and boiling light water as coolant (the main
characteristics of the reactor are shown in Table 2.3).
The RBMK reactor is based on experience with the design and many
years of operation of uranium-graphite channel-type reactors in the USSR.
Neutron physics calculation techniques which had proven themselves in
operating reactor units therefore served as the basis for developing a.
methodology for neutron physics calculations for the RBMK reactor. Two main
stages in the theoretical reactor physics studies can be identified:
(a) Calculation of the unit cell of the core and development of
constants for full-scale core calculations;
(b) Full reactor calculations taking into account the details of the
core structure.
For engineering design calculations in the first stage, use is made
of programs which make it possible to calculate the spatial energy
distribution of neutrons in a multi-group approximation in a multi-zone
cylindrical cell and also in a cell with a cluster-type arrangement of fuel
elements. For this, parameters such as the burnup of the uranium, the
isotopic composition of fuel, the power of a channel as a function of time,
the reaction rates of the isotopes of which the cell is composed and other
characteristics are determined. The bulk of the calculations are performed
for a one-dimensional cell with parameters averaged over the height.
Constants for calculations using the reactor program are prepared in the form
of a polynomial dependence on burnup and power for different average coolant
densities over the reactor height.
In the second stage, full reactor calculations are performed which
take into account the distribution of burnup over core channels, the actual
positions of control and protection system rods and the actual power of the
device. Mass calculations of states are carried out using a two-dimensional
two-group program taking into account the actual field distribution over the
height of the reactor obtained from sensors at different heights. Where
necessary, a three-dimensional program is used for performing reactor
calculations.
In addition to theoretical studies, in the design of the RBMK-l000
reactor and also during the process of operation of units already constructed,
considerable attention has been paid to experimental verification and
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adjustment of the theoretical methodologies adopted. To this end, RBMK
critical assemblies have been designed and put into operation which simulate
sectors of the reactor core. At present, an extensive programme of
experiments is being carried out in order to study the neutron physics
characteristics of the RBMK-1000 and RBMK-1500 cores, both on assemblies and
on operating units with these reactors.
With a view to achieving maximum fuel cycle economy, the RBMK-1000
reactors incorporate continuous on-load refuelling. Spent fuel is unloaded
and fresh fuel loaded with the reactor operating at a specified power level
using a refuelling machine. When the reactor goes into steady-state operating
regime (continuous refuelling regime), all the reactor characteristics are
stabilized and the fuel being unloaded from the core has an almost constant
burnup, the extent of which is determined by the enrichment of the make-up
fuel and also by the particular number of control rods introduced into the
core which is needed for ensuring the optimum field of power output over the
radius and height of the reactor. The reactivity excess for RBMK-1000
reactors adopted is 1.5-1.8% (30-36 manual control rods). With a 2%
enrichment in 235U in the make-up fuel, the burnup of fuel unloaded is
P = 22.3 MW.d/kg. It should be particularly borne in mind, that as a result
of structural materials with low absorption cross-sections and a coolant with
a high steam content being used, fuel being unloaded from the RBMK-1000
reactor in continuous on-load refuelling regime has a fissionable isotope
content similar to that in the wastes from enrichment plants, which
practically excludes any need for it to be reprocessed for recycling.
In the design of the RBMK-1000 reactor, particular attention has been
paid to demonstrating the viability of the fuel and channel element.
The main parameters determining the limiting thermal load of the fuel
channel and element are the critical channel power Neer, at which
departure-from-nucleate boiling occurs on the surface of fuel elements,
causing overheating of the fuel cladding, and secondly, the maximum
permissible linear load to the fuel element gicr, above which the dioxide
fuel melts.
In order to estimate anticipated Nc and q1 values in the reactor,
a probabilistic method for determining possible deviations was used which
takes into account different factors influencing the limiting values of Ne
and In, including the accuracy of measurement and maintenance of reactor
power as a whole and its distribution over the core (the coefficients of
inhomogeneity over the core radius Cr and height Cz and also over fuel
elements in an assembly Cass determining the maximum theoretical channel
power Mcmax and linear load to the fuel element climax). A Gaussian
distribution was assumed for random deviations in maximum power from the most
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probable value of Relax. The limiting channel power is determined from
the relationship:
Nclim = Ncmax (1 4
where oc is the mean-square error in determination and maintenance of the
channel power.
,
1 In accordance with the Gaussian distribution curve, the probability
\
of a channel with maximum power (a channel loaded with fresh fuel in the
plateau zone) exceeding a power of Nclim will be equal to
(1 - 0.9987) =0.0013.
Similarly, the limiting linear load to the fuel element is:
gliim = qlmax (1+
where
where oq, is the mean-square error in determination and maintenance of the
linear power of the fuel element. On the basis of calculations and-operating
experience, the following starting quantities have been taken for estimating
Nclim and qilim:
Cr = 1.48; Cy = 1.4; oc = 5.2%; 041 = 7.7%.
In addition to economic and thermal engineering criteria, a factor of
considerable importance - especially from the point of view of operating
safety - is the dynamic characteristics of the core. The so-called void
coefficient of reactivity o. is of particular significance. Both
experimental studies on operating RBMK units and theoretical studies show
that, with design parameters of a core in the refuelling regime adopted, the
coefficient et is positive and reaches a value of 2 x 10-6 units per
percent steam over the volume.
However, the set of means developed for controlling the RBMK reactor
includes systems which reliably ensure compensation of possible instabilities
in the power output field associated with a positive reactivity feedback in
terms of steam content. Specifically, the control and protection system
includes local automatic control and local emergency protection sub-systems.
Both operate from the signals of ionization chambers within the reactor. The
local automatic control sub-system automatically stabilizes the basic
harmonics of the radial-azimuthal distribution of power output, while the
local emergency protection sub-system provides emergency protection for the
reactor against an increase in power in individual regions of the fuel
assembly in excess of the specified level. There are 24 shortened absorber
rods for controlling the vertical fields, and these are introduced into the
core from below.
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, In addition to improvements in reactor, monitoring 4iut 'control
systems, there are other means of improving the dynamic characteristics of the
RBMK core.
These include the following:
- Increase in make-up fuel enrichment to 2.4-3.0%; leading to a
corresponding increase in burnup which makes it possible to
reduce the void effect practically to nil;
- Increase in the amount of uranium loaded into reactor channels
by using fuel compositions with a high U content.
Table 2.3 shows theoretical estimates of effects and coefficients of
reactivity associated with the variation in moderator and fuel temperatures
and also the "fast" power coefficient of reactivity.
Theory and experiments show that the "fast" power coefficient of
reactivity is negative and near zero when the reactor is operating at nominal
parameters.
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Table 2.3. Basic neutron physics characteristics of the RBMK-1000 reactor.
No. Parameters Value
1. Fuel enrichment, % 2.0
2. Mass of uranium in an assembly, kg/ass 115
3. Number/diameter of fuel elements in sub-assemblies, mm 18/13.6
4. Burnup, MW.d/kg 20
5. Coefficient of inhomogeneity of power output over the
core radius 1.48
6. Coefficient of inhomogeneity over the core height 1.4
7. Limiting theoretical channel power, kW 3250
8. Isotopic composition of fuel unloaded, kg/t
235u
4.5
236u
2.4
239Pu
2.6
240pu 1.8
241pu 0.5
9. Void coefficient of reactivity at operating point,
10-6% steam 2.0
10. "Fast" power coefficient of reactivity 4:03w, 10-6/MW -0.5
11. Temperature coefficient of fuel otT, 10-5/0C -1.2
12. Temperature coefficient of graphite ac, 10-5/0C 6
13. Minimum "weight" of control and protection system rods,
AK 10.5%
14. (Averaged) effect of replacing spent fuel by fresh fuel 0.02%
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2.5.1. Basic thermal physics data
[See end of this section for list of symbols]
2.5.1.1. Parameters determining the possibility of reactor operation in
thermal terms
For a boiling water-graphite reactor, the main parameters determining
the possibility of operation and its safety in thermal terms are as follows:
fuel element temperature, temperature of the graphite stack and the margin to
channel power at which departure-from-nucleate boiling occurs.
The condition of the hydrodynamic stability of the fuel channels of a
boiling water-graphite RBMK-1000 reactor are usually not a limiting factor
since hydrodynamic instability as a rule occurs at channel powers higher than
those at which departure-from-nucleate boiling occurs.
Experimental studies performed during design have confirmed that this
conclusion is correct and have shown that the nominal operating parameters of
the RBMK-1000 reactor are within the region of hydrodynamic stability.
If the permissible fuel temperature is exceeded or departure-from-
nucleate boiling occurs, an individual fuel assembly may become defective, but
after it has been exchanged the reactor's capacity for operation is restored.
Calculations of the margins to critical power and of the maximum fuel
element temperature in RBMK-type reactors at steady-state power levels are
performed using probabilistic statistical methods, and the same methods are
used as a basis for monitoring the core condition of such reactors during
operation.
In transient and accident regimes, when there are rapid changes in
parameters, it is advisable to assume a higher probability of the limiting
values for thermal parameters being exceeded than with operation at
steady-state power levels. Experimental data and operating experience with
boiling water-graphite reactors show that short-lived departure-from-nucleate
boiling and increase in fuel temperature above the level permitted for
steady-state regimes in these reactors do not cause fuel assemblies to become
defective.
The margin to critical power and the maximum fuel temperature in
transient and accident regimes with boiling water-graphite reactors are
determined from the actual average values for the parameters influencing these
quantities.
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2.5.1.2. Thermal physics characteristics in steady-state operating regimes
The composition of the RBMK core depends on the operating cycle.
During the first part of the operating cycle with these reactors, the core
contains channels with fuel of low burnup and a large quantity of additional
absorbers needed for compensation of excess reactivity. As burnup proceeds,
they fUel inventory in the core changes continuously. During this transitional
period-of'the cycle the core contains channels with fuel of different burnup
levels, additional absorbers of different effectiveness and also channels
filled with water. This transitional period of reactor operation ends after
all or almost all additional absorbers have been extracted from the core and
they have been replaced by fuel assemblies. The reactor is then operated in
continuous on-load-refuelling regime, in which a refuelling machine is used to
replace spent fuel assemblies by fresh ones.
A reactor operating in continuous on-load refuelling regime can be
represented as a system consisting of what can be considered
"refuelling-cycle" cells. Each "refuelling-cycle" cell consists of channels
loaded with fuel assemblies of different levels of burnup. At any given
moment, different channels will have different powers, but the total power of
all channels of a "refuelling-cycle" cell will remain approximately constant.
The design of the RBMK reactor is such that during a fuel campaign it
is possible to regulate the flow of water through fuel channels by varying the
aperture of the isolating and regulating valves at the inlet to each channel
during operation. The purpose of regulating the water flow of each channel
individually is to ensure that there is a sufficient margin to departure-
from-nucleate boiling in those core channels which are under the greatest
thermal stress, while maintaining the total flow of water through the reactor
at a moderate level. Regulation of the water flow through a channel during a
campaign is carried out in accordance with the readings of a flowmeter placed
at the inlet to each reactor channel until a theoretically determined flow
rate is reached; this is based on ensuring the necessary steam content at the
outlet from the channel or the necessary margin to departure-from-nucleate
boiling in a given channel.
A distinguishing feature of RBMK reactors is that the water flow can
be measured and controlled in each fuel channel. This ensures the
redistribution of water flow with changes in reactor power and in the power
output field over the core radius.
In accordance with the algorithm of thermal calculations for RBMK
reactors, the distribution over core fuel channels of water flow is calculated
by the standard iteration method using the combined characteristics of the
circulation pumps and the downcoming portion of the circulation circuit.
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In order to determine the hydraulic characteristics of individual
structural components and of the reactor fuel channel as a whole, experimental
studies were performed on special test rigs and also on a full-scale reactor
fuel channel simulator with a thermal power of up to 6 MW.
The relations for computing the relative coefficient of hydraulic
resistance of a cluster of rods immersed in two-phase flux have the following
form:
V
13/25 'z.1-g5-76t 1"4, " .1
x Z ? 0- 00
for the actual volume steam content in the channel:
I .4 ?
? 2C V.' A
and for the coefficient of phase transition:
K=
2
if_' 7
113;7
where Wo = G/p'S is the circulation rate and ft is the volume steam
content flow.
P
( - 22 5.7 I
_
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.Figure 2.8 compares experimental values for pressure gradients in the
heated part of the full-scale test rig and data obtained theoretically.
2
1 2 3 4
Theoretical pressure gradient, kg/co2
Fig. 2.8. Comparison of experimental and theoretical values for hydraulic
resistance of a full-scale test rig.
experimental values
theoretical values
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It will be seen from the figure that the theoretical method satisfactorily
represents the experimental data and can be used for performing thermal
calculations for reactors.
At given thermal powers in each fuel channel and at a given water
flow rate through it, the critical channel power tier, the minimum margin to
departure-from-nucleate boiling Km, the probability of departure-from-
nucleate boiling occurring in a channel R and also the probability of all core
channels operating without departure-from-nucleate boiling H are determined.
Dependences for calculating the critical power of an RBMK fuel
channel were determined following analysis and processing of experimental
information on departure-from-nucleate boiling in smooth clusters of heated
rods and in clusters of rods with heat flux intensifiers. The experimental
work was performed on test rigs with different cluster geometries (including
full size) and with coolant parameters similar to operating reactor parameters.
The dependence for calculating the critical heat flux in fuel
channels without heat flux intensifiers has the form:
3 eej3
-dh . OW 163)4574. 0,91.16-2: o(he.PW.(0-13
(0 (ticr 664?Cits7.(yM16-31i11-1-39)q 17.(E)cl,
O
4tC2cr )
9
where f(Z) is the relative distribution of power output over the channel
height;
Zer is the co-ordinate of the place in which departure from nucleate boiling
occurs (m);
and Ah is the heating of water to saturation at the inlet (kJ/kg).
The set of programs developed can be used for performing the thermal
calculations for the RBMK reactor operating in continuous on-load refuelling
regime with any positions of the isolating and regulating valves at the inlets
to each "refuelling-cycle" cell channel. By the same means it is possible to
determine the thermal parameters of the reactor with different frequencies of
individual channel flow regulation, different regulation criteria (in
accordance with outlet steam content or with the margin to critical power) and
also with core flow reduced in advance to different extents.
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The results of theoretical studies on the influence of the frequency
of individual channel flow regulation on thermal parameters of an RBMK reactor
with an electrical output of 1000 MW (RBMK-1000) operating in continuous
on-load refuelling regime are shown in Fig. 2.9. It will be seen from the
dependences shown that, when the frequency of individual channel flow
regulation, increases, the parameter H, which stands for the thermal
reliability of the core, increases; this increase is most marked when the
frequency of regulation is increased to twice per fuel campaign. A further
increase in regulation frequency does not lead to a large increase in the
parameter H. On the basis of the calculations performed during design for an
RBMK-1000 reactor operating in continuous on-load refuelling regime, it was
assumed that the water flow through each fuel channel would be regulated twice
during a fuel assembly campaign.
For the thermal calculations for a reactor operating in the
transitional period of the cycle (from the point of view of refuelling), a
mathematical model has been developed with which it is possible to derive the
distribution over the core channels of water flow and of margins to departure-
from-nucleate boiling taking into account the specific characteristics of each
individual reactor channel. In this case the reactor core is represented in
the form of a system consisting of channels loaded with fuel assemblies of
different levels of burnup and additional absorbers of any type. The
distribution over the reactor channels of power output is determined either as
a result of physics calculations for the core condition and control rod
positions being considered or is transmitted to the reactor designer by means
of a special automatic system for linking operating RBMK reactor units. As a
result of reactor calculations for the given core condition and power output
distribution over reactor channels, an optimum distribution over the channels
is found for the water flow, as is the hydraulic profile of the core needed
for this.
At operating RBMK reactors the margins to critical power and the
temperature conditions of the fuel are monitored by a special program (PRIZMA
program) using the plant's own computer. The temperature conditions of the
graphite stack are monitored using thermocouples placed over the radius and
height of the stack.
For calculating the distribution of power output over the reactor
core, use is made of the readings of a system of physical monitoring based on
in-reactor measurements of the neutron flux over the radius and height of the
core. In addition to the readings of the system of physical monitoring, data
on the core composition, the power output of each fuel channel, the positions
of control rods and the distributions over the core channels of water flow and
the readings from coolant pressure and temperature sensors are also fed into
the plant's computer. After the computer has used the PRIZMA program to
perform calculations on a periodic basis, the operator receives information on
a digital printer in the form of core diagrams showing the type of fuel load
in the core, the positions of control rods, the arrangement of the network of
in-reactor sensors and the distributions of power, of water flow and of the
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4190
499,5-
4V96-1111 41
I/
4385' cm 4o
n om
c
4-
4980-floki; 39
4975_
38
0
;a
Frequency of flow regulation
Fig. 2.9. Reactor parameters as a function of frequency of individual
channel flow regulation:
1. Main circulation pump pressure (APpump);
2. Coolant flow rate (Cf);
3. Heat content at inlet (ii);
4. Steam content at outlet (xp);
5. Thermal reliability (H).
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margins to critical power and to maximum permissible thermal loads to fuel
elements for each fuel channel. The margins to critical power and maximum
permissible thermal loads are calculated by means of a probabilistic
statistical method taking into account errors in determination of the power
output field over the height and radius of the reactor and errors in
calculation formulae and in the accuracy of measurement and maintenance of
process parameters of the plant by monitoring and measuring instruments and
automatic systems. The plant's computer also calculates the overall thermal
power of the reactor, the distribution over separators of the flow rates of
steam-gas mixture, the integral power output, the steam content at the outlet
from each fuel channel and other parameters needed for monitoring and
controlling the plant.
When the reactor is operating at steady-state power levels, and also
during periods of power increase or decrease, the operator monitors and
controls the energy output field over the radius and height of the core using
the readings from sensors of the physical monitoring system. If the field
deviates by a certain amount from the specified value, a warning light begins
to shine on a special panel. Also a warning is triggered off if the signals
from the sensors exceed the specified absolute values of the margins to
maximum permissible thermal loads to fuel elements (K41). The operator also
monitors and controls the distribution of flow rates over the core fuel
channels. The distribution of flow rates is obtained on the basis of
calcuations by an outside computer and by means of the PRIZMA program on the
plant's computer on the basis of the distribution of margins to critical power
(Km) over the fuel channels.
The temperature conditions of the graphite stack in operating RBMK
reactors are monitored by means of thermocouples placed at the corners of
graphite blocks at various points over the volume of the stack. In addition
to direct measurements of the graphite temperature at reference points in the
stack, the PRIZMA program can be used to calculate the maximum temperature
(over the height) of the graphite in the environs of any reactor fuel
channel. The graphite temperature is found on the basis of readings from the
thermocouples and of the distribution of power output over the core volume
calculated by means of the PRIMA program.
The temperature conditions of the graphite stack in RBMK reactors are
controlled by varying the composition of the gaseous mixture in the stack
(nitrogen and helium). On the basis of operating experience with Soviet
water-graphite reactors, the maximum graphite temperature at which the stack
does not burn up in the absence of water vapour has now been found to be
7500C.
Experience with operating RBMK reactors shows that, with the existing
monitoring and control systems at these reactors, maintenance of the
temperature conditions of the fuel and the graphite and of the margins to
departure-from-nucleate boiling at lhe permissible level with steady-state
power levels does not give rise to any difficulties.
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Symbols used in Section 2.5.1 (Basic thermal physics data)
G - flow rate, kg/s;
s.- cross-sectional area, m2;
d - diameter, in;
g - acceleration in gravitational field, m/s2;
- mass steam content;
p - density, kg/m3;
P - pressure, kgs/cm3
q - thermal flux density, kW/m2;
w - velocity, m/s.
Superscripts and subscripts
he - heated;
? g - hydraulic;
cr - critical;
/ - water on saturation line;
// - steam on saturation line.
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2.6. TECHNOLOGICAL LAYOUT OF THE UNIT
The technological layout consists of a single loop designed according
to the "twin unit" principle, i.e. one reactor + two turbines with no steam
and feedwater cross-connections.
At power, the unit operates in accordance with the following scheme
(see Fig 2.1.0).
The coolant (water) of the multiple forced circulation circuit (MFCC)
(primary coolant circuit) is pumped through downcomers (325 x 16 mm) from the
lower: part of the steam separators at a temperature of 265?C and a pressure
of 69 kgf/cm2 to the intake header (1026 x 63) of the main circulation
pumps. These pumps feed the water to the pressure header (1046 x 73) and then
via pipes (325 x 16) to the 22 distributing group headers of the reactor.
From the distributing group headers, the water is delivered individually to
the reactor fuel channels through the pipes of the lower water communication
lines (57 mm in diameter).
The steam-water mixture formed in the reactor passes through the pipes
of the steam-water communication lines (76 mm in diameter) and is distributed
to four steam separators in order to produce saturated steam to operate the
turbines.
Steam is removed from the top part of each separator through 14 steam
discharge pipes (325 x 19) to two steam headers (426 x 24 in diameter) which
then link up in a single header (630.x 25).
Live steam is supplied via four pipes (630 x 25) to the turbines in the
machine room (two pipes per turbine).
The pipe section located before the turbine main steam valves contains
various steam discharge devices: eight main safety valves with a throughput
of 725 t of steam per hour, four turbine condenser fast-acting steam dump
stations with a capacity of 725 t of steam per hour (two per turbine plant)
and six service load fast-acting steam dump stations. The purpose and mode of
operation of these devices are described in Section 2.7.
The exhaust steam from the turbines is condensed in the condensers.
After being completely purified in the desalination plant, the condensate from
the condensers is pumped through the low pressure heaters to the deaerator at
7.6 atm (two deaerators per turbine). Five electric feedpumps (one of which
is a back-up) -deliver the feedwater at a temperature of 1650C from the
deaerators to the steam separators where it is mixed with the circulating
coolant.
In addition to the MFCC, the reactor's main process systems include:
r-,
- - Emergency core cooling system;
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Chemical composition of primary coolant circuit materials
Material
tin
Si
SIP
Steel I 4?),23
0,9-
0,2
0,025
-(0,025
"Krezelso"
1,2
0,4
330E
Steel 0,05
40,2
40,7
i0,035
ICL47316
Material
Ni
0,3
..... ...... OMNI
Cr
(0,4
Cu
NS
(0,3
8,5-
10,5
20
Physical and mechanical properties of the materials
At tempera?
ture, ?C
// E ,
0012
Steel
"Krezelso"
330E
Steel
1CL473Nb
08Cr18NilOT1
60,2
k gf Allt2 g f ./11042 %
20 11,1 205200 44-60
350 15,0 188000 >36
20! 16,5 193000 >50
350 17,5 179000 36
20 J ,\ 16,4 205000 52
350 17,6 175000 42
8x%
C-
0265
tb
???? 0111.
0,6
. >22 >20 )48 )7 )4
49 )18 >43 )3
')20 >38 50
I5 A24 )40'
22 35 55 -
17 26 51
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A.
B.
FOR OFFICIAL USE ONLY
MAIN THERMAL LAYOUT OF THE UNIT
To pressure suppression pool
To turbine condensers
To turbine condensers
0, E, F. Original illegible
G. from ECCS
Stea? fro? turbine
I. From ECCS
J. Stea? fro? turbine
K. To pressure suppression pool
L. To station internal load
To turbine condensers
Fro. turbine condensers
O. Stea? fro? de-aerators
Original illegible
Stea? from evaporators
R. To turbine condensers
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3:-Ji
:r
e:5=-1->i7tH
Lo.o.a-Crow?
I
99
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1,
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1
- Primary cooling circuit flushing and shutdown cooling system;
- Gas circuit;
- Fuel cladding failure detection system;
Cooling pond water cooling and purification system;
- Cooling system for the water of the system "D" biological shield
tanks;
Cooling system for the control and protection system channels;
Intermediate loop of the reactor section;
Channel tube failure detection system.
Primary coolant circuit
The purpose of the MFCC is to provide the reactor fuel channels with a
continuous supply of coolant which removes the heat produced by the reactor
and to generate a mixture of steam and water which is then separated to
produce saturated steam to work the turbines. The circuit consists of two
identical independent loops each of which cools one half of the reactor. All
the equipment of these loops is arranged symmetrically about the transverse
axis of the reactor. Each circulating loop contains:
- Two steam separators;
- Water and steam connection lines between the steam separators;
- Downcomers;
Intake header;
- Main circulation pump intake pipes;
- Four main circulation pumps (three operating, one back-up);
Main circulation pump discharge pipes and fittings;
- Pressure header;
- Connection line between the main circulation pump intake and
pressure headers and their fittings;
-
Distributing group headers;
- Pipes of the lower water communication lines;
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.Reactor fuel channels;
Steam-water communication line pipes.
The downcomers, distributing group headers and the pipes of the water
and steam-water communication lines are made of stainless steel 08Crl8NilOTi.
The pressure and intake headers and pipes of the main circulation pumps are
made of carbon steel 330 E surfaced with 1CL473Nb steel from a French firm
Creusot-Loire.
The pump discharge pipes contain, in series, a check valve, throttle
valve, gate valve with remotely controlled electric drive and an orifice
meter. The presence of the gate valves on the intake and discharge pipes of
the pumps enables a pump to be removed for repair while the circuit is in
operation.
The throttle valve makes it possible to keep the main circulation pump
capacity within the unit's steady state operating range of 5500-12000 m3/h
in transient conditions. The intake and pressure headers are linked by a
connection line 750 mm in diameter, the purpose of which is to ensure natural
circulation in the loop when the pumps are not operating.' The connection line
has a check valve which prevents the medium from flowing from the pressure
header to the intake header under normal loop operating conditions, as well as
a gate valve which is normally open under all operating conditions.
Inserted in the discharge nozzles of the pressure header are leak
limiters in the case of pipe rupture. During the pre-start-up flushing
period, mechanical filters were mounted on these. The pipes supplying water
to the distributing group headers have manual gate valves. Under normal
conditions these valves are locked open; they only shut when repair work is
being done on the primary circuit. The distributing group header is fitted
with check valves, beyond which (in terms of the direction of flow) the pipes
of the water communication lines deliver water individually from the headers
to the reactor fuel channels.
Primary coolant circuit flushing and shutdown cooling system
The purpose of this system is:
- Under rated conditions, to cool the flushing water of the MFCC
before it is purified, heated and returned to the MFCC;
- Under shutdown cooling conditions
coolant circuit;
to remove heat from the primary
- Under start-up conditions, to cool the flushing water of the
circuit before it is purified, heated and returned to the circuit,
and to discharge disbalance waters from the circuit when it is
being heated.
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The flushing and shutdown cooling system comprises a regenerative heat
exchanger, large and small flushing aftercoolers, two shutdown cooling pumps,
pipes and fittings.
2.6.1. Gas circuit. Condenser and filter station
In order to prevent oxidization of the graphite and to improve the
transmission of heat from the graphite to the fuel channel, the gaps between
the graphite blocks and rings of the reactor stack are filled with a mixture
of nitrogen and helium (20 vol.% N2 and 80 vol.% He). Impurities are
removed and the nitrogen-helium ratio of the gaseous mixture is maintained by
the helium purification station.
Under normal conditions the gas circuit system functions in the
following manner. The nitrogen-helium mixture from the station passes through
the channel tube failure detection system where channel-by-channel monitoring
of the temperature and group monitoring of the humidity of the mixture being
pumped through is performed.
The mixture then enters the condenser and filter station.
The purpose of the gas circuit condenser and filter station is to
condense steam which gets into the nitrogen-helium mixture when the reactor
channels lose their leak tightness and to remove iodine vapour from the gas
mixture.
This system is designed in accordance with the principle of 100%
redundancy, i.e. it has two independent subsystems one of which functions and
the other is a back-up. Each subsystem contains a gas circuit condenser, an
electric heater and a filtration column.
The nitrogen-helium mixture from the reactor space enters the
condenser. The condensate from the condenser is removed through a water seal
to the floor drain tanks via a permanently open repair valve.
Service water is supplied to the condenser at a pressure exceeding that
of the steam-gas mixture both in rated and accident conditions.
After the condenser stage, the gas mixture has a humidity of about
100%. If it goes straight to the filter it is possible that the moisture will
condense with the result that the filter will break down. For this reason the
gas mixture is dried in the "electric heater section before proceeding to the
filtration column. The column purifies the mixture of solid particles and
iodine in aerosol form.
Ttly--filtration column is designed to purify 1000 m3 of the gas
mixture per hour._ Upon leaving the filtration column, the gas mixture goes
either to the coMpressor intake header of the helium purification plant or to
the gas activity reduction system, depending on the operating mode of the gas
circuit. 102-,
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The two sub-systems are housed in separate isolated compartments which
enables repair work to be done on the equipment of one sub-system while the
other is in operation.
2.6.2. Channel tube failure detection system
This system is equipped with sensors to monitor the integrity of the
fuel channels. Its purpose is to:
- Perform group monitoring of the humidity of the gas removed from
the graphite stack of the reactor and pumped through the system;
- Identify damaged reactor channels;
- Prevent moisture spreading from a damaged channel to adjacent
cells;
- Dry the reactor graphite stack.
At an operating reactor, channel integrity is monitored by measuring
the temperature of the gas pumped through the gaps between the channels and
the graphite stack (ducts). As the amount of steam in the pumped gas
increases, so does its temperature which is established by thermocouples
mounted in the group valves. The graphite stack with the channels which
penetrate it is conventionally divided into 26 zones, each of which contains
up to 81-channels.
The impulse tubes of the channel ducts in each zone run to the
corresponding group valve for that zone. Each of the 26 group valves is
designated by the same number as its corresponding reactor zone.
The valve outlet nozzles of the are connected by pipes to the channel
tube failure detection system ventilation and intensified extraction headers.
Both these headers are linked to the process gas circuit, thus joining the
channel tube failure detection system to the reactor's process ventilation
circuit.
It is possible to alter the gas pumping rate through the impulse tubes
leading to a given valve by switching the slide valve between the ventilation
system and the intensified extraction system.
2.6.3. Helium purification plant
The purpose of the helium purification plant is to purify the gas
mixture circulating through the RBMK unit closed circuit of oxygen, hydrogen,
ammonia, steam, carbon oxide, carbon dioxide, methane and nitrogen impurities
to a level which permits the reactor to operate normally.
The gas mixture becomes contaminated because moisture can enter the
stack cavity as a result of the non-leaktightness of the fuel channels. The
moisture then partially decomposes as a result of radiolysis into hydrogen and
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oxygen which, reacting with carbon, forms carbon oxide and carbon dioxide. The
hydrogen which combines with graphite forms methane and that with nitrogen,
ammonia.
The main technical characteristics of the helium purification plant are:
1. Mixture quantity at 293 K and 101325 Pa 0.0833-0.264
(760 mmHg), m3/s(m3/h) (300-950)
2. Pressure at plant inlet, MPa (mm water) 0.003 (300)
3. Composition of unpurified mixture, % (vol.):
nitrogen 20
oxygen 0.3
methane 0.1
ammonia 0.07
carbon dioxide 0.02
carbon oxide 0.1
hydrogen 0.6
chlorine traces
helium residue
4. Pressure at plant outlet, MPa (mm water) 0.005 (500)
5. Temperature at plant outlet, K (?C) 308 + 10 (35 + 10)
6. Composition of purified mixture, % (vol.):
nitrogen 10
oxygen 0.01
methane traces
ammonia traces
carbon dioxide and oxide 0.01
hydrogen 0.02
helium residue
7.
Auxiliary products used:
liquid nitrogen, kg/s (kg/h)
gaseous nitrogen, m3/s (m3/h)
gaseous oxygen, m3/s (m3/h)
cooling water, m3/s
0.039
0.097
0.0042
0.0056
(140)
+ 0.104 (350-500)
(15)
(20)
8.
Duration of operating run, years
1.5
9.
Duration of startup period (s/h)
57
600
(16)
10.
Time preceding first overhaul, h (year)
43
800
(5)
11:
Life expectancy, year
30
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2.6.4. Cooling and purification system for spent fuel cooling pond water
The purpose of this system is to sustain the temperature regime of the
fuel assembly and pressure tube cooling pond water, which is heated by the
afterheat of the spent fuel and pressure tubes. The system is designed in
accordance with the principle of 1007. redundancy and comprises pumps, heat
exchangers, pipes and fittings. The function of the pumping station of the
purification plant is to deliver the cooling pond water to the ion-exchange
filters of the purification plant. There is one such water purification unit
for two power units. It operates periodically.
2.6.5. Cooling system for biological shield tanks
The purpose of the pumping-heat exchanger station is to control the
temperature of the water in the reactor biological shield tanks. It contains
the following equipment: cooling circuit circulation pumps, heat exchanger,
expansion tank, pipes and fittings.
2.6.6. Cooling system for control and protection system channel
Function and design bases
The function of the cooling system for the control and protection
system, fission chamber and power density monitoring channels as well as the
reflector cooling channels is to ensure that the prescribed temperature is
maintained in these channels. The system must meet the following requirements:
Maintain the prescribed temperature in the aforementioned channels
in all operating regimes of the unit (startup, power operation,
shutdown, disruption of normal operating conditions, accident
situations);
Satisfy the regulations regarding for water quality (chemical
composition and specific activity).
The system comprises a circulation loop which operates by gravity feed;
in other words, the water flows through the channels as a result of the
difference in levels of the upper and lower tanks.
Water from the upper tank (known as the emergency storage tank) passes
through a pipe (400 in diameter) to the pressure header and is distributed
among the channels.
The volume of the emergency tank is governed by the condition that it
should supply the rated flow through the channels for six minutes when the
pumps are out of operation.
The cooling water from the pressure header enters the channel from
above, passes down through the central tube, and then travels up out of the
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channel through the annular gap between the central and outer tubes to the
discharge header of the reflector cooling channel. There are two discharge
headers (200 in diameter).
The water from the discharge header moves along the pipe (400 in
diameter) to the system's heat exchangers. Into the same pipe flows water
from the reflector cooling channel discharge headers: before entering the
400 diameter pipe, the water from these headers converges in a common pipe 150
in diameter. This common pipe is equipped with a throttle device which
eliminates the need for a syphon in the reflector cooling channel discharge
headers.
There are six heat exchangers in the system to cool. the circulating
water when it leaves the reactor.
Having passed through the heat exchangers, the water enters the
circulation tank through a pipe (400 in diameter) below the water level. In
the circulation tank the flow is slowed down and conditions are created for
hydrogen to be separated efficiently from the water. A description of the
method used to keep the hydrogen concentration at a safe level in the tank is
given in Section 2.7.
Water from the emergency storage tank is continuously dumped into the
circulation tank via overflow pipes. The amount corresponds to the difference
between what the pumps deliver and the throughput of the control and
protection system, fission chamber, power density monitoring and reflector
cooling channels. If two of the system's pumps are working, an overflow pipe
150 in diameter is used; if three pumps are working, a pipe 300 in diameter is
used.
The system is equipped with four pumps to feed water from the
circulation tank to the emergency storage tank. Two of these operate and the
other two are back-up. The first back-up pump is switched on automatically,
while the second is activated by the operator when required.
The pumps receive power, from a category 1B secure supply from the
diesel generators.
,In order to maintain the required water quality in the circuit, the
water is continuously purified by a bypass system at a rate of 10 m3/h.
2.6.7. Intermediate circuit system of the reactor section
The objective of the reactor section intermediate circuit is to prevent
radioactive substances .from getting into the service water from the heat
exchangers of systems containing radioactive coolant should these lose their
leaktightness. This is achieved by keeping the pressure in the intermediate
circuit lower than that of the service water.
The circuit is a closed system consisting of an expansion tank, pumps,
and heat exchangers as well as shut-off, safety and regulating components.
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The circuit pumps supply cooling water to the heat exchangers of the reactor
section systems and remove heat fromthem; this in turn is absorbed by the
intermediate circuit heat exchangers which are cooled by service water. An
expansion tank is used to keep the pumps operating smoothly and to prime,
supply make-up and to compensate for changes in the volume of intermediate
circuit coolant. With regard to auxiliary reactor systems which are located
at higher levels and for which cooling water cannot be supplied by the main
circulation pumps, of the circuit, higher-pressure pumps have been installed
to deliver water to the heat exchangers of the steam separator sampler and to
the reactor refuelling machine.
Periodic or continuous cleaning of the water in the reactor section
intermediate circuit by special purification plants is not necessary. The
quality of the intermediate circuit water is determined by sampling. When
increases in the chloride content or changes in the pH value of the medium
exceed the established limits, the water of the intermediate circuit is
purified by exchanging the water in the system.
Systems using the intermediate circuit are the reactor flushing and
shutdown cooling system, the equipment controlled leakage system, the main
circulation pump sealing water coolers, the helium purification plant and heat
exchangers of the chemical monitoring sampler.
2.6.8. Water regime
The reliability, safety and economics of fuel element operation and
normal radiation conditions at a nuclear power plant are governed by the
water-chemical conditions of the main and auxiliary circuits. The
water-chemical conditions of these systems must satisfy the following main
requirements:
- Reduce the amount of contamination getting into the reactor core;
- Prevent the deposition on core components of impurities contained
in the water.
A neutral water regime is used in RBMKs whereby radiolysis of the water
is not inhibited and no additives are introduced to correct the pH.
In accordance with All-Union Standard 95743-79, the quality of the
coolant in the MFCC should meet the following requirements:
- pH value: 6.5-8.0;
- Specific conductance:
not greater than 1.0 pohm/cm;
Hardness: not more than 10 pg-equiv./kg;
- Silicon acid:
not more than 100 pg/kg;
- Chloride-ions + fluoride-ions:
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not more than 100 pg/kg;
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- Iron corrosion products: not more than 100 pg/kg;
- Copper corrosion products: not more than 20 pg/kg;
- Oxygen: 0.05-0.1 mg/kg;
Oil: not more than 200 pg/kg.
The'quality of feedwater must meet the following requirements:
- pH value: 7.0;
Specific conductance: not more than 0.1 pohm/cm;
- Iron corrosion products: not more than 10 pg/kg;
Oxygen: 0.03 mg/kg;
During plant operation, the water-chemical regime prescribed for the
circuit must be permanently maintained and radioactive water purified before
being reused and discharged. Radioactive water at nuclear power plants goes
to an active water treatment system consisting of a number of plants. These
plants can be divided into main and auxiliary categories.
The main active water treatment plants include the following:
- Bypass cleaning of the primary coolant circuit flushing water;
- Cleaning of cooling pond water;
- Cleaning of cooling water for the control and protection system;
Cleaning floor drains;
Cleaning the controlled leakage system;
- Cleaning wish-out and resin regenerating water;
- Cleaning primary coolant circuit decontaminating solutions;
Cleaning the pressure suppression pool water.
Auxiliary plants of the active water treatment system include the
following:
- Preparation of regenerating solutions;
- Perlite preparation and deposition;
Filter loading;
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- Transfer of resin to the solid and liquid waste storage tank;
- Preparation of decontaminating solutions;
- Reuse of decontaminating solutions;
- Equipment decontamination.
Apart from the MFCC flushing water bypass purification plant and the
pressure suppression pool water purification plant, the installations listed
above are housed in block B on axes 35-41 at levels 0.00, 6.00 and 12.50 and
serve two units.
The MFCC flushing water purification bypass plants are located in
block A and block B. The pressure suppression pool water purification plant
and the floor drain mechanical filter preliminary purification plant are
housed in the radioactivity treatment auxiliary system block.
2.6.9. Bypass purification plant for MFCC flushing water
The purpose of this plant is to purify in the bypass mode primary
circuit flushing water in order to remove corrosion products and dissolved
salts. This system is the main means of maintaining the quality of the
circuit water, preventing deposits forming on fuel elements and ensuring the
long working life of the MFCC. It enables non-volatile fission radioisotopes
to be removed from the circuit, induced activity to be reduced and, most
important of all, radioactive contamination of the steam and
condensate-feedwater channels to be reduced. Each unit has its own
independent system of this type.
The system is designed to purify 200 t of circuit water per hour. This
capacity is governed by the flushing rate with regard to corrosion products
and enables the circuit water characteristics stipulated in the regulations to
be maintained. Under steady-state conditions the capacity of the system may
even be lower. Under transient conditions at a pressure not exceeding
16 kgf/cm2, the MFCC decontaminating solution purification plant can be used
to remove corrosion products which have accumulated under steady-state
conditions. This enables the rated value for iron corrosion product flushing
in the MFCC to be maintained during reactor startup and shutdown cooling.
The system components are:
1. One mechanical ion exchange filter;
2. Two mixed-bed ion exchange filters;
3. One filter trap;
4. One moisture trap.
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2.7. Main equipment of the unit
Reactor
The series-produced power reactor RBMK-1000 is used as the plant's
steam-generating system. The reactor and its technical characteristics are
described in Section 2.2.
Turbine
The K-500-65/3000 fast turbine with underground condensers is used as
the mechanical drive for the AC generator TVV-500-2UZ.
The principal rated characteristics of the turbine unit are given in
the table on the next page.
Steam separator
The steam separator of the RBMK-1000 is intended for obtaining dry
saturated steam from the steam-and-water mixture.
The separator is a horizontal cylindrical vessel with elliptical
bottoms having 400 mm openings.
The steam-and-water mixture comes to the separator through 632 short
pipes of the steam-water communication line, which are located in the
cylindrical part of the lower half of the separator in four rows on each
side. The kinetic energy of steam-and-water mixture is quenched and crude
separation of the steam takes place at the baffles inside the separator.
Thereafter, the steam passing through the immersed plate is separated
in the steam space and, passing through the perforated ceiling plate, leaves
by way of 14 short pipes located in the upper generatrix of the separator.
In the body of each separator there are four nipples for monitoring
steam pressure and 24 nipples for connection of water gauges.
The separator is mounted on five supports, the middle one of which is
fixed, while the others are of the sliding guide type.
The major assembly parts and components of the steam separator are made
of the following materials:
(a) Body and bottom: 330E steel + 1C 473 B (clad steel), made by
Creusot-Loire, France (for composition and properties, see
Section 2);
(b) Short outlet pipes for steam: 330E steel;
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No. Characteristic
Unit Value
1. Maximum turbine power, MW MW 550
2. Net rated turbine power BMW 510
3. Rated fresh steam flow, including
fresh steam for the second stage
of the steam heater t/h 2890
4. Maximum fresh steam flow, indluding
fresh steam for the second stage of
of the steam heater
5. Initial steam pressure
6. Initial steam temperature
7. Initial moisture content of steam
8. Feed water heating temperature
9. Rated pressure in the condenser
10. Type of steam distribution
t/h
kgf/cm2, abs.
Oc
PC
kgf/cm2, Abs.
reducer type
11. Turbine design diagram 2 LP cylinder + HP
cylinder + 2 LP cylinder
12. Structural formula of steam
regeneration diagram 5 LP heater + de-aerator
13. Number of steam sampling
regeneration
14. Frequency of rotation rev/min
15. Turbine load for heating (intermediate
circuit graph 160/80?C) Gcal/h
Technical characteristics of condenser
16. Quantity of steam condensed (per
condenser)
17. Cooling water temperature at
condenser inlet
18. Number of passes of cooling water
19. Cooling area
20. Condenser hydraulic resistance
2902
65.9
280.4
0.5
168
0.05
7
3000
75
t/h
441.105
oc
18
2
m2
12150
m of water column
3.63
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(c) Short inlet pipes for the mixture and short downcomer pipes for
circuit water: clad steel - 330E steel + 1C 473 B;
(d) Separator internals: 1C 473 B steel.
The technical data on the steam separator are:
- Steam output, t/h 1450
- Saturated steam pressure, kgf/cm2:
working
rated
70
75
Moisture content of steam at
separator outlet, .% not exceeding 0.1
- Steam temperature, ?C
Feed water pressure at steam
separator inlet, kgf/cm2
Feed water temperature,
- Flow of circuit water, t/h
- Flow of steam-and-water
mixture, t/h 9400
284.5
71
165
9400
Average steam content in steam-and-
water mixture entering the separator, % not exceeding 15.4
Accuracy of level control in the
steam separator in relation to
rated value, mm not exceeding + 50
Effective water margin in the separator
separator for a possible level of
100 mm below the rated value, m3 not less than 51
Separator service life, years 30
- Weight of the steam separator: dry, t 280
in working state, t
during hydraulic test, t
394
439
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- Basic dimensions of the separator:
length, mm 30 984
inner diameter of body, mm 2600
minimum wall thickness of main
metal, mm 110
De-aerator
The de-aeration unit consists of a de-aerator tank and two de-aeration
columns. The tank has three supports: the outer two are of the roller
sliding type, which enable the de-aerator to expand during heating, and the
middle one is fixed, which restricts displacement of the central part of the
de-aerator in the horizontal plane and allows movement in the vertical plane.
The working pressure is 6.6 kgf/cm2 and temperature 167.5?C. The weight
of the de-aerator during hydraulic tests (completely filled with water) is
204 t.
Main circulation pump
It is a centrifugal vertical single-stage pump. The shaft has a
double-end packing with insignificant supply of sealing water, precluding
escape of coolant into the building.
The main characteristics of the pump are:
- Delivery 8000 moP/h
- Head 200 m of water column
- Temperature of coolant pumped 270?C
- Pressure at pump suction opening 72 kgf/cm2
- Minimum permissible cavitation margin 23 m
- Pump shaft power 4300 kW
Electric motor power 5500 kW
The unit consists of a tank, a pumping part and an electric motor.
The tank, which is a welded structure, is made of the 15Kh2MFA Icr2mov]
steel and has an anti-corrosion surfacing inside. It constitutes the support
of the pumping part and is connected to the latter through a joint which is
hermetically sealed with a packing. The pumping part contains the shaft with
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the working wheel, the distributor, a lower hydrostatic bearing, an end
bearing and an upper thrust guide bearing, which are located in the housing.
The pump is so designed that the pumping part can be partly or fully replaced.
Water is supplied to the lower hydrostatic bearing from the general
pressure header of pumps through a hydrocyclone.
The thrust guide bearing has a circulating lubrication system with
filtration and cooling of oil from the auxiliary oil system of each pump.
The pump allows prolonged operation at a rate in the range of
5500-12000 m3/h. The permissible heating end cooling rate of the pump is
20C/mm.
Feed pump
An electric pump unit is used for supply of feed water from the
de-aerators to the steam separators.
The electric pump unit is a three-stage pump with the working wheels
located on one side and with a preset screw, hydraulic journal, slit-type end
packings and sliding bearings with forced lubrication. Cold condensate
(t = 400C) is supplied to the pump packing.
The content of mechanical impurities in the condensate should not
exceed that of the feed water in weight and volume.
The main characteristics of the unit are:
- Delivery 1650 m3/h
- Head 84 kgf/cm2
- Feed water temperature 169?C
Pressure at pump suction opening 9 kgf/cm2
Minimum permissible cavitation margin 15 m of water column
- Pump shaft power 4200 kW
- Industrial water flow 36.5 m3/h
- Oil flow 3.5 m3h
- Cold condensate flow 21 m3h
- Electric motor power 5000 kW
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Condensate pumps
The condensate is supplied from the condensers to the de-aerators
through a low-pressure heating system by means of condensate pumps of pressure
rise I and II.
The condensate pump of pressure rise I is a centrifugal vertical
double-barrel electric pump with a preset wheel and end packing of
interchangeable types: gland and end-face types.
The basic characteristics of this unit. are:
- Delivery 1500 m3/h
- Head 12 kgf/cm2
- Pressure at suction opening,
not more than 0.2 kgf/cm2
Condensate temperature
up to 60?C
- Minimum permissible cavitation
margin, not less than 2.3 m of water column
- Pump shaft power 615 kW
- Condensate flow to the end-face
packing 3 m3/h
Flow of cut/lints water LO
pump bearings 1.5 m3/h
- Electric motor power 1000 kW
Weight of the unit 24 560 kg
The condensate pump of pressure rise II is a centrifugal horizontal spiral
type electric pump unit with a working wheel of two-sided entry. The end
packings are of two interchangeable types:
- End-face packing: for continuous operation;
- Gland packing: for trial startup operations.
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The basic characteristics of the unit are:
- Delivery
- Head
Pressure at suction opening,
not more than
1500 m3/h
240 kgf/cm2
15 kgf/cm2
Condensate temperature up to 60?C
Minimum permissible cavitation
? margin, not less than
- Pump shaft power
Electric pump power
Weight of the unit
Pipelines
22 m of water column
1150 kW
1600 kW
10335 kg
The pressure and suction headers (average diameter 800) of the multiple
forced circulation loop (MFCL) and main circulation pump inlet and outlet
pipes (average diameter 800) are made of the 330E carbon steel with a
surfacing of the 1C 473 B steel supplied by the firm Creusot-Loire (France).
?The MFCL pipes with a diameter of 300 mm are of the 08khl8N1OT
[crl8nilOti) stainless steel. .The pipelines of the reactor auxiliary systems
are of. carbon steel. The condensate feed channel pipes are of steel 20.
Fresh-steam pities are made of the 17 GS steel.
Refuelling machine
A most important requirement which the RBMK reactor must satisfy is
that it should operate with a minimum number of shutdowns. For this reason,
there is provision for refuelling and control of some accident situations in
the operating reactor without reduction of power. This is ensured by a
special refuelling machine, which carries out the following operations:
Loading and unloading of fuel assemblies in the operating and the
cooled reactor;
Verification of free passage through the fuel channel using a
gauge simulating a standard assembly;
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- Hermetic sealing of the fuel channel
emergency type);
with a plug (normal or
- Mechanized control of some accident situations.
Refuelling in the operating reactor is carried out at the working
parameters of the fuel channel.
In 24 hours the refuelling machine can carry out five operations of
fuel channel unloading and loading in an operating reactor without reduction
in its power and not less than 10 such operations in a shutdown reactor.
The principal parts of the machine are a crane, a container, two
detachable pressure equalization chambers (one at the machine and the other in
the repair area), a frame, technological equipment, a guidance system and
control organs.
The working principle of the machine in the operating reactor is
described below.
The refuelling machine filled with condensate at 30?C is attached to
the channel to be refuelled. A pressure equal to that in the fuel channel is
created in the pressure equalization chamber and the channel is unsealed. The
condensate is pumped from the pressure equalization chamber into the channel
at a rate of up to 1 m3/h. The cold condensate prevents the penetration of
steam and hot water from the channel into the refuelling machine. After
removal of the spent fuel assembly the channel is sealed and the pressure in
the pressure equalization chamber is reduced to atmospheric pressure. The
machine is disconnected from the channel and sent to the location where spent
fuel assemblies are unloaded.
The refuelling machine has two systems of accurate guidance into the
fuel channel: optical-TV (main) and contact (stand-by) in case of loss of
visibility in the steaming channel.
The optical-TV system allows visual observation of the image of the end
of the channel head on a TV screen or through the eye-piece of this system and
alignment of the circle of the channel head with the dotted circle of the
sight by small movements of the bridge and the carriage. The contact guidance
system is a pneumatic-electromechanical device, which guides the refuelling
machine on to the channel axis by means of direct mechanical contact /of the
system with the lateral surface of the channel head.
The refuelling machine is controlled from the operator's cabin, which
is behind the end wall of the central hall.
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In addition, the refuelling machine cabin has a control panel for crane
movement.
The central hall includes the following service areas for the machine:
1. Parking site: an area in the central hall intended for parking
the machine in the period between refuelling operations.
2. A simulator rig intended for:
Adjustment and checking of the machine's mechanisms;
Filling of the pressure equalization chamber with
condensate;
Simulation of regular refuelling;
Loading of fresh fuel
equalization chamber;
assemblies into the pressure
Decontamination of the inner space of the pressure
equalization chamber;
Replacement of the inflatable collars
sleeve.
the connecting
The simulator rig has the appropriate equipment for these
operations.
3. The facility that receives spent fuel assemblies is used for
keeping the gauge.
4. The repair area is intended for replacement of the pressure
equalization chamber if it is out of order. This area is situated
in the central hall in the region of the simulator rig. A fully
assembled spare pressure equalization chamber is always available
in the area.
The equipment of the safety systems is described in
Section 2.9.
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2.8. Control and protection ustem
The control and protection system (CPS) of the ?RBMK reactor provides:
- Control of the level of the neutron-flux-determined power of the
reactor and its period under all operating regimes from
8 x 10-12 to 1.2 Nnom;
- Startup of the reactor from the shutdown state to the required
power level;
- Automatic regulating of the reactor. power at the required level
and changes in that level;
- Manual (from the operator's control desk) regulating of the power
density distribution throughout the core and regulating of the
reactivity to compensate for burnup, reflection and other effects;
- Automatic stabilization of the radial-azimuthal power density
distribution in the reactor;
- Preventive protection - rapid controlled reduction of the reactor
power to safe levels: protection level 1 = 50% Nnom , protection
level 2 . 60% Nnom;
- Emergency protection when the parameters of the reactor or
generating unit change as a result of an accident (protection
level 5).
The CPS (for a structural diagram see Fig. 2.11) comprises:
- Neutron flux sensors
in the reactor;
with devices (hangers) for positioning them
- Reactivity regulating devices (absorber rods) with drive
mechanisms which move the regulating rods within the reactor
channels;
- The equipment of the CPS measurement subsystem, which converts the
information from the neutron flux detectors and generates discrete
signals for subsequent processing in the CPS logic subsystem, as
well as analog signals for the indication and recording of reactor
parameters;
- The equipment of the CPS logic subsystem, which carries out the
prescribed control and protection algorithms; the CPS logic system
processes discrete signals from the CPS measurement and drive
subsystems, from the command devices at the control desk, from the
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unit's automatic process systems and from other systems; the
result of this processing is the generation of a command: to move
the control rods under normal and emergency conditions, to change
the power level, to change the operating regimes, or to give
signals;
- The equipment of the CPS drive subsystem, which controls the servo
drive mechanisms in accordance with the commands from the CPS
logic system;
Output devices for indicating and recording reactor and CPS
parameters at the control desk and instrumentation board;
The CPS electrical power supply system.
2.8.1. Location of the main CPS equipment items
The hangers for the neutron flux detectors are located:
- In the tank for the water shielding around the reactor, where
there are 24 ionization chamber hangers; of these, 16 have KNK-53M
working-range ionization chambers and 8 have KNK-56 startup-range
ionization chambers;
- During startup, 4 hangers with KNT-31 fission chambers are lowered
into the reflector channels; after assured monitoring by the
startup chambers has been achieved, the fission chamber hangers
are withdrawn from the reflector;
- In the central openings of the fuel assemblies there are
24 in-core detectors with KTV-17 fission chambers.
All 211 CPS drive mechanisms are mounted above the CPS channels in the
reactor. Their servo drives are of the channel-mounted type. The position of
the CPS rods is indicated by means of a selsyn transmitter installed in the
servo drive mechanism and a selsyn receiver (rod position indicator) on the
CPS mimetic diagram panel in the operator's control and instrumentation
board. The extreme positions of the rods are determined by cut-off switches,
installed in the servo drives, which actuate the cut-off upper- and lower-end
lights in the corresponding position indicators.
,The equipment of the CPS drive subsystem is based in the CPS location
_behind4he central hallwall; it consists of:
The servo drive control panel for the manual regulating and
emergency protection system and the shortened absorber rods;
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- The panel for the automatic regulator servo drive control blocks,
consisting of three boards with individual servo drive power
control blocks;
- The control rack for the local automatic regulating system servo
drives (model BA-86), containing 12 sections;
- The servo drive temperature monitoring rack.
Three boards with the automatic regulating rod synchronization system
blocks are installed in the location of the non-operative part of the unit
switchboard.
The CPS measurement subsystem equipment is located in the non-operative
part of the unit switchboard and consists of various electronic instruments
installed in 19 panels of the electronic instrumentation board, and of 2
plug-in racks holding the electronic instruments of the local automatic
regulating and local emergency protection systems.
The system of output indicators and automatic recorders is installed in
the operator's control desk and instrumentation board.
The CPS logic subsystem equipment is also housed in the non-operative
part of the unit switchboard.
Signalling by the CPS system, which is accompanied by auditory signal
and flashing lights on the signal board located in the control and
instrumentation board, is carried out by devices in the CPS cupboard in the
non-operative part of the unit switchboard.
- The command devices (keys, buttons, etc.) which the reactor operator
uses to control the CPS rods, change the reactor power, switch operating
regimes, etc., are accommodated in the control desk.
2.8.2. Neutron flux monitoring
The monitoring ranges are shown in Fig. 2.12.
Neutron flux monitoring in startup regimes over the range 8 x 10-12 -
3 x 10- 7Nnom is performed by four independent measurement channels with
KNT-31 fission chambers. The sensitivity of the chamber to neutron flux is
0.25 pulses/l/cm2. The secondary electronics (ISS.3M counting rate meters
with KV.3M output cascades) operating from the fission chambers determine the
neutron flux density on a logarithmic scale and the reactor excursion period.
The information output from these channels is displayed on indicators at the
control desk and recorded from a single channel selected by the operator.
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At intermediate power levels in the range 3 x 10-8 - 5 x 10- 2Nnom
the neutron flux is monitored on the basis of signals from four KNK-56 startup
current ionization chambers with an enhanced sensitivity to a neutron flux of
4 x 1013 All (cm2.$)-1. To reduce interference from the gamma
background the chamber channels are surrounded by lead screens. Additional
compensation for the gamma background is achieved by regulating the negative
feed voltage of the chambers' compensating electrodes. The signals from these
chambers, with their secondary electronics (UZS.13 protection system amplifier
with KV.2 logarithmic output cascade), determine the neutron flux density on a
logarithmic scale and the reactor excursion period and generate signals to
reduce the excursion period to the alarm and emergency settings (alarm signals
and emergency protection system). The information output from these channels
is displayed on indicators at the control desk and recorded on tape from one
of the channels at the control and instrumentation board.
The discrete signals from the alarm and emergency protection systems
concerning the reactor excursion period are processed in the 'protection
system's logic circuit.
Neutron flux monitoring and recording on a linear scale in the range
8 x 10-8 - 1.0 Nnom is also performed by 2 KNK-53M ionization chambers
with a neutron flux sensitivity of 1.45 x 10- 14 A/I (cm2.$)-1. The
secondary device in this case is a KSPV 4 high-impedance multi-range recorder.
The reactivity is measured by a ZRTA-01 reactimeter with 10 reactivity
measurement ranges from 0.01 to 5 R. The reactimeter monitors the neutron
flux (power) of the reactor, which is displayed on an indicator in the control
desk with a scale selector and recorded by a special unit in the control
board. The channel with the reactimeter operates on the signals from 2
KNK-53M ionization chambers.
2.8.3. Automatic regulating of the reactor power
The system comprises three identical sets of automatic regulators for
the average reactor power. Each set consists of four ionization chambers
placed around the reactor and providing information on the basis of which four
automatic regulating rods are moved synchronously. The automatic regulating
signal is generated by summing the relative deviations of the power from the
required level, which are determined in the four individual ionization chamber
measurement channels. This design principle ensures that the automatic
regulator will remain functional when one ionization chamber or the
instruments in one measurement channel fail.
The equipment in all three automatic regulator sets is identical.
The use of ionization chambers of different sensitivity enables these
sets to work in different ranges: the low-power range from 0.5 to 10% Nnom
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and the working-power range from 5 to 1001. Nnom. In the low-power range
there is one automatic regulator (3AR); in the working-power range there are
two (1AR and 2AR).
The detector and part of each measurement channel of the automatic
regulator are also used as a power overshoot protection channel: four power
protection system channels in the low-power range and eight channels in the
working-power range.
A structural diagram of the CPS is given in Fig. 2.11.
The detector signal in each channel is corrected by a KrT.5 current
corrector. The corrected signal is compared with a reference signal from the
Zd.M.5 power transducer which is common to each set of four channels. The
unbalance signal is transmitted to the UZM.11 power protection amplifier and
the US0.10 deviation signal amplifier. When the unbalance signal reaches the
value set in the alarm and emergency protection systems for a power overshoot,
the UZM.11 amplifier generates alarm and emergency signals, respectively, for
further processing in the emergency protection system logic circuit. The
US0.10 amplifier, whose gain can be regulated by the power transducer,
produces a signal indicating relative deviation of the power from the required
level. Information on the power deviation in the places monitored by the
detectors is displayed on the unbalance indicator at the control desk and to
some extent allows the operator to monitor the power density distortions
throughout the reactor. The output signals from the US0.10 amplifiers of the
four channels are summed in the USM.12 amplifier, which then transmiti
information about the deviation of the average power from the required level
for display on the indicator at the control desk which shows when the
automatic regulator is switched on. From the output terminal of the summing
amplifier the signal is transmitted to the automatic regulating rods
synchronization system, which synchronizes the positions of the automatic
regulating rods. This synchronization system generates a relay law for power
regulation. It also produces a signal indicating the average rod position for
a given automatic regulator and signals indicating the deviation of the
individual rods' position from the average. On the basis of the signal for
the relative deviation of the average power from the required level (from the
USM.12 amplifier output terminal) and of the signals for the deviation of the
rod position from the average, a command is formed for the withdrawal or
insertion of the automatic regulating rods into the core. These signals
control the automatic regulating rod servo drives via BKS.40 power control
blocks.
One of the working-range regulators is switched on, while the second is
in "hot" standby. The second regulator is automatically switched on if the
first regulator is switched off automatically as a result of a malfunction.
In order to switch the standby regulator on smoothly, i.e. without moving the
automatic regulating rods, a zero unbalance is automatically maintained at the
output terminal of its summing amplifier by means of a KrU.4 automatic
corrector.
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The CPS ensures that identical settings are obtained from the power
transducers in the working range with an accuracy no worse than 0.57. Nnom.
The transducer settings are synchronized by a BSP.36 block and a logic circuit
on the principle of stopping the transducer which has the leading setting in
the direction of change of the settings.
The transducer settings are controlled by the operator from a key at
the ,control desk. The operational rate of change in the transducer settings
does not exceed:
0.0075% Nnom per second in the range 0.5-1% Nnom;
0.0125% Nnom per second in the range 1-6% Nnom;
0.15% Nnom per second in the range 5-20% Nnom;
0.25% Nnom per second in the range 20-100% Nnom.
Under emergency conditions the settings of the working transducer
settings are automatically reduced at a rate of 2% Nnom per second. The
settings can also be lowered in an emergency by means of a button in the
control desk.
The automatic regulators give a power holding accuracy for the reactor
no worse than + 1% in relation to the required level in the range
20-100% Nnom and no worse than + 3% in the range 3.5-20% Nnom.
In addition to the monitoring of correct functioning which can be
carried., out on the various blocks of the system, there is also a continuous
automatic monitoring of the correct functioning of the working-range automatic
regulator measurement channels, including the neutron flux detectors. The
BT.37 block compares the output signals of the analog channels with the
signals of the channels from neighbouring detectors around the reactor. When
a channel signal deviates from both its neighbours by an amount exceeding the
actually possible distortions in the reactor, the channel in question is
regarded by the circuit as being out of order. This type of monitoring is
used at steady-state power levels and is automatically switched off in
emergency conditions and transitional regimes of the generating units.
When the working-range automatic regulators are operating, the 3AR rods
may be brought on-line for overcompensation of the automatic regulator which
is switched ,on. In this case, when the rcds of the switched-on automatic
regulator emerge as far as the intermediate cut-off switch, corresponding to
75-100% rod insertion, the 3AR rods are automatically moved downwards, but at
the intermediate cut-off switch corresponding to 25-0% rod insertion they are
moved upwards.
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Stabilization of the power density distribution in the reactor is
achieved by the local automatic regulating and local emergency protection
systems. The former is designed on the principle of independent power
regulating in 12 local zones of the reactor by means of 12 regulating rods.
The local automatic regulating system rods are controlled on the basis of
information from two KTV.17 chambers positioned in the core around the local
automatic regulating rods at a distance of 0.63 m from the rods.
The KTV.17 chamber is a current ionization chamber whose sensitive
elements are coated with a 235U compound and which incorporates a guard
electrode to reduce loss of useful signal. The BP.119 power supply block
places a negative voltage on the collecting electrode. The guard electrode
receives a voltage of the same magnitude and polarity as the central
collecting electrode; thus, both the guard and the collecting electrode are
under an identical potential and the current losses from the collecting
electrode are reduced to a minimum. The KTV.17 chamber has three sensitive
elements arrayed along the height of the core.
The local automatic regulating system is switched into the automatic
mode in the power range after the information received from the power density
physical control system has indicated that the required power density
distribution has been achieved. Before it is switched on, the output signals
from the local system zones are compensated by means of the system's
correction devices. Then the system, while holding the power value set before
switching on in each of the 12 zones, stabilizes the power distribution in the
reactor. The overall power is held by the local automatic regulating system
with an accuracy no lower than that of the traditional average-power automatic
regulating system. In transitional regimes, the local automatic regulating
system also has considerable advantages, since it not only provides
measurement and regulation of the overall power, but also smoothes out power
distortions due to local perturbations in the equipment.
The local automatic regulating system is now the main system for
automatic power regulating in the power range from 10 to 100% Nnom. The
average-power automatic regulating system is used for standby and is
automatically switched on when the local automatic regulating system is
switched off is a result of malfunction.
The local automatic regulating system, consisting of 12 physically
independent local regulators, has a high degree of "viability": when several
zones of the system are switched off or malfunction, the system as a whole
remains operational.
The signal from each chamber is corrected by a KT current corrector.
After passing the corrector, part of the signal is transmitted to the local
emergency protection system channel, where alarm and emergency signals
indicating power overshoots over the required level are generated; part of the
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signal from each of the two chambers in the local automatic regulator zone is
summed in the deviation signal amplifier, which generates the signal
Indicating relative deviation of the power in the local automatic regulator
zone from the required level. When the values given by this unbalance are
exceeded, the trigger Tg puts out signals to move the local automatic
regulating rods in the corresponding zone. The speed of movement of these
rods is reduced to 0.2 m/s so as not to exceed the limits laid down by the
Nuclear Safety Regulations for the rate of insertion of positive reactivity
when 12 rods of the local system are moved at the same time (0.7 POWs).
There is a built-in limitation on the continuous withdrawal of the
automatic regulator rods for over eight seconds.
When a power overshoot alarm signal appears in one of the channels of
the local emergency protection zone, the withdrawal of the local automatic
regulating rods is automatically blocked. When emergency power overshoot
signals appear in both channels of the local emergency protection zone, two
local emergency protection rods are lowered into this zone of the core until
at least one of the emergency signals disappears. In this case the average
power of the reactor is reduced by automatic lowering of the power transducer
settings at their operational rate change.
The withdrawal of more than 8-10 of the manual regulating and emergency
protection system or shortened absorber rods upon any malfunction (in the
control desk, CPS logic, servo drive power control blocks etc.) is prevented
by the "power blocking" circuit. This circuit automatically determines the
number of rods in whose servo drive armature circuit a voltage for rod
withdrawal is given. If this number is greater than 8-10, the circuit is
automatically disconnected from the servo drive power supply source, and not a
single rod can be withdrawn from the core. There are three power blocking
channels which process the signals by a two-out-of-three logic.
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2.8.4. Emergency protection of the reactor
The reactor is protected against emergencies by the automatic insertion
into the core of all absorber rods (except for the shortened rods) from
whatever initial position along the height of the core.
Twenty-four CPS rods uniformly distributed through the reactor are
selected for the emergency protection mode from the total number of manual
regulating and emergency protection rods by a special selector circuit
installed in the CPS logic racks. When the reactor is started up, the 24
emergency protection rods are the first to be raised to the upper cut-off
switches; the withdrawal from the core of. any other rods is automatically
prevented until the emergency system rods have been raised; the arrival in the
raised position of the selected emergency protection rods is automatically
verified and notified.
The reliability of the emergency protection system and the reliable
functioning of the manual control system is achieved by effectively having six
independent groups of 30-36 control rods each distributed uniformly through
the reactor. Each CPS rod is moved by its own servo drive under the control
of its individual power and logic block. The rods are connected in their six
Rroups by the layout of the servo drive power supply and control blocks and by
the design layout of the control blocks. The failure of one or even several
servo drives or control blocks is not serious, since their total number
is 187. Generalized reasons for the failure of several independent groups are
ruled out. Since each CPS rod is surrounded in the reactor by rods of
different groups, the failed rod is always surrounded by neighbouring rods in
working order.
The design of the CPS drive mechanisms is such as to ensure automatic
insertion of all CPS rods (except the shortened rods) into the core in a power
failure. The reliability of the protection system is ensured by functional
redundancy (redundant monitoring channels) for each parameter and equipment
redundancy (redundant channels for logical processing of the signals).
In view of the large contribution of nuclear power plants with RBMK
reactors to the general power grid, it is necessary to reduce to a minimum the
outages of such plants; a differentiated approach to emergency situations in
the reactor and generating unit has therefore been adopted in organizing the
emergency protection system. Depending on the nature of the emergency
situation, there are a number of different categories (regimes) for emergency
protection:
- Emergency protection with complete shutdown of, the reactor -
protection level 5;
- Emergency protection acting until the emergency situation has
passed - protection level 5*;
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Preventive controlled reduction of reactor power at an increased
speed to safe levels: protection levels 3, 2 and 1; the safe power
levels for various emergency situations and the speed of preventive
power reduction are determined by calculation and confirmed
experimentally.
The highest level of emergency protection is level 5, which is achieved
by inserting all the CPS rods (except the shortened absorber rods) into the
core up to the lower cut-off switches. This regime is entered in the
following situations:
- A power overshoot of 10% Nnom;
A reduction in the period to 10 s;
A drop or excess in the level in the drum separators of either half;
A drop in the feedwater throughput;
- A pressure excess in the drum separators of either half;
- A pressure excess in the leaktight compartments, drum separators or
lower water lines;
- A pressure excess in the reactor cavity;
A fall in the level in the CPS coolant tank;
- A reduction in water throughput through the CPS channels;
Trip of two turbogenerators, or of the only operating
turbogenerator;
- Trip of three of the four operating main circulation pumps in any
pump room;
Voltage loss in the plant auxiliary power supply system, or
indication of one of the protection level regimes (3, 2 or 1)
without its being carried out, or order from the command units
(protection level 5 button, declutching key) at the control desk
and at a number of other locations in the plant.
, In the event of an emergency power overshoot (power protection system)
detected by the lateral ionization chambers, a partial alarm regime described
as "protection level 5*" is ordered in which the insertion of the CPS rods
into the core is interrupted when the original cause of the emergency has
disappeared (when the power has been reduced to the appropriate level). This
makestjt possible to keep the unit in a power regime if the power overshoot
signals have been caused by power distortions and the emergency situation can
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be removed by rapid partial reduction of the overall reactor power. The same
is true in transitional operating regimes of the unit and in the case of
significant local perturbations. The protection level 5* regime can only
operate for a short time, for if the CPS rods are lowered to a significant
extent into the core during a protection level 5* event, the reactor will be
completely shut down just as in a normal protection level 5 regime.
The protection level 3 regime is ordered when there is an emergency
load rejection by two turbogenerators, or by the only operating one.
The protection level 2 regime (reduction of N to 50%) is ordered in the
following situations:
Outage of one of two turbogenerators;
Emergency load rejection of one of two turbogenerators.
The protection level 1 regime (lowering of N to 60%) is ordered when:
One of the three operating main circulation pumps in any pump room
is switched off;
When the water throughput in the primary circuit falls;
When the feedwater throughput falls;
When the water level in the drum separators falls;
When the group closure
actuated.
key for the throttle regulating valves is
In protection level 1, 2 and 3 regimes the reactor power is
automatically reduced at a rate of 2% Nnom/s to levels of 60%, 50% and 20%,
respectively, by the on-line automatic power regulating system. The emergency
rate (speed) of power reduction and reactor operation stabilization at a safe
power level after its reduction are obtained by automatic switching into the
automatic regulating regime of the supplementary CPS rods (overcompensation
and protection system) - the overcompensation regime. Signals initiating the
protection level 1, 2 and 3 regimes, as well as the level 5 regime, are
carried out for technical reasons in the automatic equipment system.
The generation of an emergency signal with respect to any parameter
occurs upon the response of two or more detectors out of the four installed.
The logic part of the emergency protection system is designed for technical
reasons as two independent sets of equipment. In order to cut the protection
system out during testing, there are individual keys for each parameter. When
the protection is switched in, this is signalled and recorded by the "Skala"
centralized control system. The protection system also has provision for
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signals indicating the actuation of the protection system, signals for the
Initial cause triggering the protection system and signals showing
malfunctions in the protection system equipment.
A structural diagram of the emergency protection system for process
parameters is shown in Fig. 2.13.
Both ordering the protection level 3, 2 and 1 regimes when the reactor
power exceeds the safe level for such situations and carrying out the
executive algorithm for levels 5, 3, 2 and 1 are the responsibility of the CPS
logic circuit.
Reliability of the protection against exceeding the speed of power
increase (the speed protection system) and of that against reactor power
overshoot (the power protection system) is ensured in the setting-generating
system for the protection level 5 regime by both functional redundancy
(presence of not less than 3 monitoring channels with their sensors for each
parameter) and equipment redundancy (logical processing of discrete signals by
several independent channels in parallel).
A protection level 5 regime relating to speed protection is ordered
when the reactor excursion period decreases to 10 s, as detected by not less
than two channels out of three:
By the startup-range speed protection system from 4 x 10-7 to
5 x 10-2 Nnom;
By the working-range speed protection system from 10-5 to
1.2 Nnom.
Each channel of the speed protection system consists of a UZS.13 speed
protection amplifier with a KV.2 logarithmic output cascade and a KNK-56
current ionization chamber with a lead screen on the channel where it is
mounted (in the startup range), and a KNK-53M current ionization chamber (in
the working range).
A protection level 5 regime relating to power protection is ordered:
By the low-power power protection system in the range from 0.005 to
0.1 Nnom, when a power level given by the ZM power transducer is
exceeded by 0.005 Nnom (by 0.5% Nnom), as detected by not less
than two out of four of the low-power protection system channels;
By the working-range power protection system in the range from 0.06
to 1.2 Nnom, when the set power level is exceeded by 0.1 Nnom
(10% Nnom)--,---as detected by two out of the eight working-range
power protection system channels; in this case there must be an
emergency protection signal in at least one channel of each of the
two groups of four working range channels.
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Each working-range power protection system channel consists of:
- A UZM.11 amplifier (power protection system);
- An ionization chamber: KNK-56 in the low power range and KNK-53M
in the working range;
- A BP.39 ionization chamber power supply block;
A KrT.5 chamber current corrector.
Each group of four working-range channels has a common ZM.5 power
transducer; one transducer for the low-power range and two transducers for the
working range. The ionization chamber., the chamber power supply block, the
chamber current corrector and the power transducer at the same time. form part
of the automatic regulator measurement channel of the corresponding range:
The presence of eight power protection channels in the power range with
transducers distributed uniformly around the core, in conjunction with the
protection system against overall power overshoot, allows the reactor to be
monitored and protected against local power overshoots.
A coincidence circuit for the signals from the two independent groups
(of four working-range channels each) with alternating detector locations
reduces the probability of false (unjustified) reactor shutdowns when there is
a malfunction of one channel or of the common element of a group - the power
transducer. Dangerous failures of the emergency protection system are ruled
out by the fact that the measurement and logic subsystems are designed on the
principle whereby any malfunction of a block or channel is equivalent to an
emergency protection signal in that channel. This design of the system makes
it possible to replace any block in a single channel for repairs and
preventive maintenance while the reactor is operating at power, which is
particularly important for RBMK reactors with an on-load refuelling option.
The working-range power protection system is ready to respond at all
times, whereas the action of the low-power system is blocked by the operator
by means of a key in the control desk when the operating range of the
low-power system has been passed.
A preventive power reduction is carried out by the automatic regulating
system which is on-line: local automatic regulating system or lAR or 2AR; by
means of automatic lowering of the power transducer settings through
protection level 3, 2 and 1 signals.
When the transducer setting is lowered, the measurement part of the
automatic regulator generates power deviation (unbalance) signals. Unbalance
signals for +1% from the on-line automatic regulator cause the rods in that
regulator to be moved, while +2.5% signals actuate the overcompensation and
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protection system rods in protection level 3, 2 and 1 regimes. Initially, two
groups of six rods each are lowered; when the rods in these groups reach the
lower cut-off switches in the presence of an unbalance of +2.5%, the
corresponding rods of the two next overcompensation and protection system
groups are lowered. Only one group of six rods in the overcompensation and
protection system is moved upwards. The +2.5% signals are generated in the
KrU.4 block of the average-power automatic regulator on the basis of a signal
from the summing amplifier.
If a preventive lowering power reduction in the protection level 3, 2
and 1 regimes is ordered by an on-line local automatic regulating system, then
+2% relative unbalance signals generated in the local system trigger block
cause the local emergency protection system.rods to move in the corresponding
zone of the local automatic regulating system. The withdrawal of the local
protection system rods from the zone is allowed only after withdrawal of the
local automatic regulating rods to the upper cut-off switch.
If the transducer setting of an on-line automatic regulator is not
reduced at the emergency speed, or if there is no on-line automatic regulator,
or if an automatic regulator has been switched off during a power reduction
without another regulator being switched on, then a level 3, 2 or 1 regime is
automatically converted into a protection level 5 regime.
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Figure 2.11:
1. Shortened absorber rods
2. Manual regulating and emergency protection rods
3. Local automatic regulating system rods
4. Automatic regulator rods
5. Servo drives
6. Servo drive control relay contact block - shortened rods
7. Servo drive relay control contact block - manual rods
8. Servo drive control block - local automatic regulating system
9. Power control block - automatic regulator system
10. Rod synchronization system.-, low power automatic regulator
- automatic regulator 1
- automatic regulator 2
11. Individual rod control circuit
12. Shortened rods manual rods local automatic regulating system
13. Command devices - control desk, control and instrumentation board,
standby control desk
14. Automatic process equipment
15. Signalling system
16. Protection level 5
17. Local emergency protection system
18. Local automatic regulating system on and in working ord
19. Power transducer setting control
20. Low-power protection system
21. Low-power automatic regulator on and in working order
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22. KNK-53M
23. KNK-56
24. KNT-31
25. KNK-53M
26. KV.2
27. KV.5M
28. BP.38
29. BP.38
ZRTA reactimeter
30. Power recorder (N)
31. Reactivity recorder (p)
32. Working-range speed protection system, channel 1
channel 2
33. BP.38
Working-range speed protection system, channel 3
UZS.13
34. Current indicator
Period indicator
35. Startup-range speed protection system, channel 1
. channel 2
channel 3
36. BP.38
TsU.1
startup-range speed protection system
UZS.13
37. Period indicator
38. Current indicator (g)
39. Power recorder (N) on logarithmic scale
40. Period indicator
41. Speed indicator
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42. BP.38M Counting rate meter, channel 1
ISS.3M
43. Counting rate meter, channel 2
channel 3
channel 4
44. BP.39
45. Power recorder (N) at standby control desk
46. Working-range speed protection system
47. Startup-range speed protection system
48. Automatic regulator 1 on and in working order
49. Power protection system
50. Automatic regulator 2 on and in working order
51. Indicators at control desk and board
52. CPS logic
53. Indicators at control desk and board
54. Power transducer setting
55. )
56. ) Unbalance, power protection amplifier
57. Unbalance, deviation signal amplifier
58. Local automatic regulating system on
59. BTE
60. Synchronized drive block
61. Trigger block
62. BTE
63. Power transducer
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64.
Synchronized drive block
65.
Local automatic regulating system, zone , 5,
7,
9,
11
66.
Local automatic regulating system, zone 2, 4, 6,
8,
10,
12
67.
Power protection amplifier
68.
Current corrector
69.
KtV.17
70.
Deviation signal amplifier
71.
Control key
72.
Power supply block
73.
Unbalance, power protection amplifier
74.
Unbalance, deviation signal amplifier
75.
Power transducer setting
76.
BT.37
77.
ZH.9
78.
Low-power automatic regulator, channel 1
US0.10
UZM.11
KrT.5
BP.38
79.
USH.12
80.
Low-power automatic regulating system, channel 2,
3,
4
81.
KNK.56, KNK.53M
82.
Automatic regulator 1, channel 1, 2, 3
83.
USH.12
84.
Unbalance, deviation signal amplifier
Unbalance, power protection amplifier
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85. Power transducer setting
86. Unbalance, deviation signal amplifier
Unbalance, power protection amplifier
87. BT.37
88. US0.10
UZM.11
KrT.5
BP.39
Automatic regulator 1, channel 4
89. KrU.4
90. ZM.9
91. BSP.36
92. KrU.4
93. ZM.9
94. Automatic regulator 2, channel 1
US0.10
UZM.11
KrT.5
BP.39
95. Automatic regulator 2, channel 2, channel 3, channel 4
96. USM.12
97. KNK.53M
98. Alarm and emergency protection systems
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Neutron flux monitoring ranges.
i
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(1) Range of power monitoring on logarithmic scale by ISS.3M units wit.
KNT-31 chambers.
(2) Range of power monitoring on logarithmic scale by UZS.li?i(spee,
protection in startup range) units with KNK-56 chambers (with lea,
screen).
(3) Range of power monitoring by low-power automatic regulator using KNK-5I
chambers.
(4) Range of power monitoring by automatic regulator 1 (2) using KNK-531
chambers.
(5) Range of power monitoring on logarithmic scale by UZS.13 (spee,
protection in working range) units with KNK-53M chambers.
(6) Range of power monitoring on linear scale by automatic recorder a,
standby control desk using KNK-53M chamber:
(7) Range of power monitoring on linear scale by automatic recordim
potentiometer, at control and instrumentation board using KNK-531
chambers.
(8) Range of power monitoring by local automatic regulator channels usint
KtV.17 chambers.
2 12
7. 7
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Pressure in leaktight compartment
Signal
multiplier
>2
Shown here is a diagram of the
protection system for increasing
the pressure in the leaktight
compartments.
The diagrams of the protection
systems for other process
parameters are analogous.
Fig. 2.13 Structural diagram of protection system for
process parameters.
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2.9. Reactor process monitoring system
The reactor process monitoring system provides the operator with
information in visual and documentary form on the values of the parameters
which define the reactor's operating regime and the condition of its
structural elements: the process channels, the control channels, reflector
cooling, graphite stack, metal structure and so on.
The following systems relate to the process monitoring system:
- Channel-by-channel coolant flow rate monitoring in the process and
control channels;
- Temperature monitoring
structure;
the graphite stack and the metal
Channel integrity monitoring from the temperature and humidity of
the surrounding gas;
Physical power density monitoring system;
- Fuel cladding failure detection;
- "Skala" central monitoring system.
Information from the PIC system is collected and processed by the
"Skala" central monitoring system, and by individual instruments or
independent systems (channel failure detection, physical power monitoring
system, fuel cladding failure detection) for the more important parameters.
The reactor has the following monitoring points:
- Fuel channel flow rate measurement: 1661 points;
- Control channel flow rate measurement : 227 points;
- Temperature measurement of the metal structure and biological
shielding: 381 points;
- Measurement of the graphite stack and plates: 46 points
- Radial and vertical power measurement: 214 points;
- Gas temperature measurement: 2044 points;
- Measurement of coolant activity: 1661 points.
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2.9.1. Coolant flow rate monitoring
Flow rates in all the reactor channels are measured using tachometrical
ball flowmeters. The flowmeters includes the primary ball sensor, a'magnetic
induction transducer and an electronic transistor unit. For measuring the
flow rates in the process channels, flowmeters are used with a range up to
SO m3/h and up to 8 m3/h in the control channels. The flowmeters in the
process channels operate in a temperature range between 20 and 2850C and
pressures up to 10 MPa, and at temperatures between 20 and 80?C and
pressures of 5 MPa in the control channels. The flowmeters are accurate to
1.5%, and have a positive systematic error -component due to temperature, which
is determined on a high-temperature flowmeter metrology rig and is
automatically corrected by the "Skala" central monitoring system. The
response of the flowmeters is six seconds or less.
The coolant flow rate in each process channel and in the control
channels is monitored by the computing unit. Channel-by-channel flow rates
are compared with norms which are set as a function of the characteristics of
the channels and their position within the reactor, and which can alter when
the operating conditions of the station change. When the computer unit
detects a breach in the limits set for coolant flow rate, it sends an error
signal to the channel mimic board and the group error mimic board, registers
on the teletype the fact that an error has appeared and blpcks the CPS system
when the water flow rate in the control channel falls below the permitted
level.
Regular diagnostic checks are carried out on the condition of the
primary sensors and magnetic induction transducers, and decisions are
regularly taken as to whether it is possible for them to continue in use or
whether prophylactic replacement should be carried out. Diagnostic checks of
the primary sensor are carried out periodically by displaying the signals from
the magnetic induction transducer on an oscilloscope, determining the signal's
amplitude and period ratio and comparing that ratio with a given criterion.
The magnetic induction transducer undergoes periodic diagnostic testing
through monitoring the resistance of its magnetic coil.
Section 2.9.2.
Temperature Monitoring
The temperatures of the graphite core and metal structures are
monitored using mass-produced Chromel-Alumel cable-type thermoelectric
transducers.
For monitoring the temperatures of the graphite stack and the upper and
lower, metal plates, an assembly of thermocouple units is used. The
thermbcouple assemblies are situated along the longitudinal and lateral axes
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of the reactor in 17 cells at the points where the corners of the graphite
blocks meet. The temperature of the graphite stack is measured using 12
three-zone assemblies (of which four are in the reflector), while five
two-zone assemblies measure the temperatures of the support and upper
shielding slabs. A thermocouple assembly contains cable heat sensors and a
support structure consisting of a biological shielding plug, graphite bushes
and connecting tubes. (Fig. 2.36.). In the three-zone assemblies, the
functional junctions of the temperature sensors are positioned in the central
cross-section of the core and at 2800 mm below and 2700 mm above the central
cross-section. The temperature sensors are manufactured from cable with an
external diameter of 4.6 mm and a caging of carbonization-resistant
high-nickel alloy. The cable has four cores with magnesia insulation and
contains two Chromel and two Alumel thermoelectrodes moulded into a single
functional junction. Each heat sensor thus contains two thermocouples with a
common functional junction.
The systematic error component in the measurement of the graphite core
temperature, caused by heat released internally within the thermocouple
assembly elements, equals 2.2% of the measured value and is allowed for by
correcting the measurement output in the "Skala" central monitoring system.
The thermal response time of the assembly is within acceptable limits
at 90 seconds, and is much faster than the thermal response time of the
graphite core, which is between 30 and 40 minutes. In the two-zone
thermocouple assembles, the functional junctions of the thermocouples are
level with the upper and lower plates.
The temperatures of the other metal structures are monitored using
cable-type Chromel-Alumel heat sensors made from thermocouple cable 4 mm in
diameter inside hermetic steel sleeves (Fig. 2.37). The sleeves are designed
not only to protect the thermocouples, but also to act as guide elements when
less accessible locations are monitored. In this way, it is also possible to
replace thermocouples which have become unserviceable. The temperature
monitoring of the metal structures is designed to determine their condition in
stationary and transient conditions. For the upper and lower metal'
structures, which are more complicated, contain a larger number of structural
elements and are acted on by significant thermal stresses, the maximum number
of control points is 30. The temperatures are monitored of the external
surfaces of the fuel channel ducts and control channels, element fins, roller
supports, expansion joints and the upper and lower plates.
The reactor casing temperature is monitored at four points along a
single vertical generatrix. The metal support structure is monitored at six
points along a single radius. For the metal structures of the upper covering
in the central hall, the temperatures of the undersides of the beam casings
are monitored (8 points). In addition, the temperature of the ,water in the
water biological shielding tanks is monitored using headed Chromel-Alumel
temperature probes (16 points) (Fig 2.38).
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_ .
The temperature of the water at the control channel discharges is
monitored using six cable-type Chromel-Alumel temperature sensors at reference
points.
One hundred and fifty-six Chromel-Copel heat sensors are used to
monitor the water temperature in the reflector cooling channels.
/
The temperature measurement equipment used has a relatively fast
response time: the thermal response time of the cable-type thermocouples is
In the order of 5 seconds, and 60 seconds when they are installed in an
additional protective sleeve for measuring the temperatures of load-bearing
metal structures. Instrument error is in the order of 2% of the temperature
measurement range.
Temperature information is periodically printed by the "Skala" central
monitoring system, and it is also possible to call up any of the parameters
using the call-up feature on the numerical display unit of the "Skala" central
monitoring system and on the redundant instrument set.
2.9.3. Process channel integrity monitoring
The process channel integrity monitoring system is part of the reactor
ventilation system and, in general, is designed to carry out the following
functions:
Detection of non-hermetic reactor channels;
Containment of the spread of humidity from the damaged channel;
Ventilation of the reactor space.
The main outlines of the process channel integrity monitoring system
are shown in Fig. 2.3.9. The process channel integrity monitoring system for
RBMK reactors relies on measuring the parameters of the gas (temperature and
humidity) as it is pumped round the graphite stack of the reactor through the
gas du ts formed by the stack and the process channels. In this way, there is
indivi ual monitoring of the temperature of the gas drawn off and group
monito ing of its humidity. The gas circulates through the graphite core from
botto to top. Temperatures are measured by short Chromel-Copel heat sensors
insta led at each process channel integrity monitoring impulse tube.
Tempe ature signal information from the thermocouples is passed for processing
to t e "Skala" system which, when it has detected the channel (or group of
chan els) with a temperature overshoot, sends a signal to the channel mimic
board on the reactor unit's control panels.
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Humidity monitoring in each of the 26 zones and detection of zones
where there is excess humidity are carried out -using humidity indicators. A
humidity indicator consists of eight humidity sensors and one eight-channel
humidity measuring unit. The sensing element of the humidity sensor is of the
sorption type and is designed to operate at temperatures of between 40 and
1000C and relative humidities from 50 to 100%. The relative humidity unit
gives readings in steps of 5%. When the humidity indicator operates, it sends
signals to the "Skala" system which are reflected on the humidity board on the
reactor unit control panel. The relative humidity of the gas in the reactor
space is continuously monitored within a range of 0 to 100% by means of a
hygrometer, which consists of a sorption-type primary sensor, measuring unit
and recorder. In order to increase the reliability of the system for
determining process and CPS channel integrity, there is a system for drainin4
the sylphon bellows cavities of the CPS channels and measuring the
temperatures of the drainage pipes. The humidity which appears in the reactor
space when a loss of integrity occurs evaporates And, as it condenses on the
nearest "cold" control channels, settles in. part' in the sylphon bellows cavity
and from there passes to the lower part of the duct into the drain pipe. When
this occurs, the temperature of the drain pipe, which passes through the lower
water communication line housing and is at the same temperature when there is
no flow, decreases, and this is picked up by a thermocouple. The region in
which the search for the burst is carried out is defined by the thermocouple
readings, which give temperature values approximately 1000C below the
temperature of the lower water communication line housing. The temperatures
of 126 control channel drainage ducts are measured and processed through the
"Skala" central monitoring system for periodic print-out.
2.9.4. Fuel element cladding integrity monitoring
The physics and design characteristics of an RBMK power station -
channel-type reactor with boiling coolant - determine the structure of the
system for identifying and locating fuel assemblies with burst fuel elements
within the core while the reactor is in operation. The physical system for
monitoring fuel cladding integrity comprises:
- A sampling system for monitoring the activity of gaseous fission
products in the separated steam in each drum separator; this makes
it possible for the condition of the fuel elements in a quarter of
the fuel assemblies in the core to be observed continuously;
- A non-sampling channel-by-channel system for periodically
monitoring total gamma activity of the coolant in each steam-water
communication line, the secondary part of which is electronic and
compensates for the background component of the signal in order to
distinguish the gamma activity of fission products emanating from
burst fuel elements.
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2.9.5. Monitoring of the multiple forced circulation circuit
The physical monitoring system for the primary coolant (multiple forced
circulation) circuit is designed to ascertain the condition and operating
modes of its basic elements: the drum separators, main circulation pumps and
the suction and pressure headers. It includes monitoring of drum separator
level and pressure, drum separator metal temperature, equalizing tank
temperatures, main circulating pump flow rates and drum separator steam out
and feed water in flow rates.
Platinum resistance thermometers are used in the suction headers to
measure the coolant temperature in order to determine the cavitation margin.
Pressure is monitored in the main circulation pump suction and pressure
headers. The flow rate through the main circulating pumps is measured using a
differential manometer, for which the pressure drop is created on the
constriction principle. The primary coolant circuit parameters are monitored
by the "Skala" central monitoring system.
2.9.6. The "Skala" central monitoring system
The "Skala" computerized central monitoring system is designed to carry
out monitoring of the processes in the basic equipment of RBMK-1000 nuclear
power station units, and to provide calculations and logic analysis of the
units' process conditions in finished form for the operating staff. A diagram
of the "Skala" system's structure and of its links with external systems (CPS,
processed channel integrity monitoring system, physical monitoring system and
so on) is shown in Fig. 2.40. The basis of the system is a two-processor
computing unit which is designed to be able to capture information from the
source and transmit it to the output devices using either of the two
processors (functional back-up). Information on the condition of the unit
coming from the process monitoring system sensors through the individual
signalling channels or through the computer unit and is passed by the operator
to the display and digital instruments, the mimic diagram, channel mimic board
and the individual error board and is also registered on recorders, teletypes
and high-speed printers. The information the operator needs to work is
accessed in the "Skala" system by means of a number of input-output devices.
The operation of the system as a whole is organized by the system control
unit. The basic technical features of the "Skala" system are as follows:
(1) Number of monitoring signals:
Analogue signals
Discrete signals
Signals are accepted from:
7200
6500
Chromel-aluminium [Alumel?] and Chromel-Copel heat sensors;
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Platinum and copper resistance thermometers;
Tachymetric ball flowmeters;
Sensors: diffusion manometers with standardized outputs of 0-5 mA,
selsyns, on-off sensors, independent physical power monitoring
system and the average control rod position signals.
(2) Monitoring periods:
Mass parameters: 1-5 min.
Calculated parameters: 30 min.
(3) Functions:
Measurement of the parameters input through the group and
individual information capture channels, and also, when commanded
by the staff, on the group, individual and digital display
instruments.
Signalling on the mimic diagram, processe channel mimic board,
group error board and CPS mimic board of the conditions of
mechanisms, fittings, generator sets, process parameters and
correct equipment function.
Monitoring of directly measured errors and errors in calculated
parameters with results shown on output devices and also recorded.
Process calculations on a periodic basis and on request.
Print-out of any of the measured and calculated process parameters
on a periodic basis and on request, with record made of the run-up
to and development of accident situations.
(4) Operation time to failure:
Monitoring functions: 1 x 104 hours;
Calculating functions: 2 x 103 hours.
(5) Electric power required: 95 kW.
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2.9.7. System for physical monitoring of the power density distribution
? Purpose and structure of the system
This,system is intended to measure and record signals from the power
density monitors which characterize the energy release in the reactor. By
initially, processing the signals coming in from the monitors and then
comparing them with preset maximum values, the system makes recommendations to
the reactor operator for regulation of the power density distribution. The
light, and sound signals emitted by the system are used for operative
flattening of the power density distribution. For additional correction of
the density distribution use is made of the link between the physical
monitoring system and the "Skala" computer system, where on the basis of the
signals from the monitors, results of the physical calculation and other
requisite information there is periodic calculation and recording of the power
and the maximum permissible power margin for each fuel assembly, as well as
calculation of other parameters for different assemblies and the reactor as a
whole.
For operative monitoring of the thermal power of the reactor between
the minimum verifiable level and the nominal level, use is made of the
monitoring system's automatically recording potentiometer, which records the
total monitor current over the reactor radius and has a scale graduated in
megawatts (0-4000 MW). A doubling instrument is used for the same purpose.
According to its functional purpose the system for physical monitoring
of the power density distribution is divided into three systems: a system for
physical monitoring of the radial power distribution, a system for physical
monitoring of the vertical power distribution, and an auxiliary system for
periodically checking the monitors.
The radial power density distribution monitoring system is intended for
measurement and recording of signals from 130 in-core detectors monitoring the
power density over the reactor radius, for preliminary processing of the
signals, for transmitting them to the "Skala" computer system, for comparing
the ;signals with three set levels and emitting light and sound signals
indicating that the power density values in the fuel assemblies fitted with
monitors have overshot the prescribed limits. The maximum power of the
assemblies with radial monitors is determined by the "Skala" system computer
on the basis of the requirement of flattening the power density distribution
and ensuring the safety of the given and neighbouring assemblies.
The vertical monitoring system is designed to measure and record
sisnals from 12 in-core seven-section power monitors in a vertical direction,
for preliminary processing of the signals, transmission of them to the "Skala"
computer, comparison of the signals with three set values and emission of
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light and sound signals indicating that the local power density in
neighbouring assemblies with monitors has overshot the prescribed limits. The
maximum values of the signals from different sections of the vertical monitor
are determined by the "Skala" computer on the basis of the requirement that
there should be stabilization of the axial power density distributions and
safe operation of the assemblies without overshooting the maximum local
thermal loads.
Periodic checking of the monitors is intended for routine calibration
of the sensitivity of the radial and vertical monitors, as well as for
determining the error involved in calculating the fuel assembly power with the
"Skala" computer system.
The system for physical monitoring of the radial power density
distribution includes:
130 detectors for monitoring the radial power density of the reactor;
measuring devices of the power density monitoring equipment;
automatic recording potentiometer (system power recorder) and doubling
indicating instrument;
scanning instruments of the power monitoring equipment.
The system for physical monitoring of the vertical power density
distribution includes:
12 seven-section detectors for monitoring the vertical power density of
the reactor;
measuring device of the power density monitoring equipment;
scanning instruments of the monitoring equipment.
The measuring devices have a block-type structure and are serviced by a
common multichannel recorder belonging to the power monitoring equipment,
which records on a numerical printing device the detector signals exceeding
the relevant maximum levels, time of occurrence and the detector co-ordinates.
The layout of the radial and vertical monitors and the CPS control rods
ensuring monitoring and control of the power density distribution in the
reactor is shown in Fig. 2.1.a. For purposes of monitoring and controlling
the power distribution the reactor design makes provision for the arrangement
of - 310 suspensions and assemblies with a dry central supporting tube
(casing). Of these 130 are intended to hold the radial monitors and 48 of
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them to hold the local emergency protection and control systems, and at least
130 remain free (assemblies for scanning the power) and are used for periodic
calibration of the radial monitor sensitivity. Assemblies of this type are
placed alongside the assemblies fitted with radial monitors.
The measuring devices and multichannel recorder are positioned in the
non-operative part of the unit control panel. The devices for light signals
indicating that the radial and vertical signals have deviated from the norm
are located on the CPS-physical monitoring system mimetic board on the
operator's control panel. The unit that switches on the light signals on the
mimetic board in response to commands arriving from the measuring devices of
the power monitoring equipment is part of the "Skala" system.
The power recorder of the power density physical monitoring system is
located on the reactor operator's control panel, while the scanning instrument
doubling its readings is on the control desk. The scanning instruments of the
power monitoring equipment are located on the control panel.
The monitor verification system includes:
calibration detectors of the radial power monitor type;
tri-axial calibration fission chambers;
annular ionization chamber;
measuring equipment of the monitor verification system.
The installation and removal of the calibration detectors from the
reactor is remotely controlled by means of the crane in the central hall. The
measuring equipment is situated in the central hall or in the crane operator's
room in the central hall.
Radial power density monitors
The radial power density monitors are enclosed in dry central
supporting zirconium cases with an internal diameter in the core of 6.5 mm and
arranged along the axis of the assembly (all the way along). The design of
the radial monitor is shown in Fig. 2.41. It consists of a sensitive element
in a leaktight casing (4) made of corrosion-resistant steel with an outside
diameter of 6 mm; a leaktight join (2), a cable (3) inside a leaktight
protective sheath, and the elements of the biological shielding (1). The
casing is filled with inert gas (argon) to protect the envelope of the
sensitive element against corrosion.
The overall length of the monitor is 16 167 mm, and the length of the
sensitive element is 8500 mm.
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As the sensitive element, use is made of a beta emission detector with
a silver emitter (5). It takes the form of a high-temperature cable
(type KDMS(S)) with an outer diameter of 3-nun, central silver filament 0.65 mm
in diameter, envelope made of corrosion-resistant steel and insulation made of
0.8 mm thick magnesium oxide. The cable is manufactured using the industrial
technology normally adopted for high-temperature cables and thermocouples.
Its sensitivity is - 5 x 10- 2? A.cm2.s/n per metre of length. The
maximum current of the radial monitor at nominal reactor power is about
15*pA. The maximum temperature of the sensitive element as a result of
radiation heat-up is greater than the coolant temperature in the assembly and
amounts to - 3500C.
The mean ratio between the power of a non-spent assembly with a radial
monitor and the current of a non-spent radial monitor is 0.2 MW/pA.
Variations in this ratio for each radial monitor due to its individual
sensitivity and neutron spectrum are taken into account by periodically
calibrating the monitor during operation of the reactor. The mean square
spread of the radial monitor's sensitivity to the neutron flux is, according
to experimental data, 4%. At the same time, the mean square spread of the
sensitivity to the power of the fuel assembly is greater and amounts to 6%,
which is explained by the difference in neutron spectrum in the different
assemblies with radial monitors. This effect may be taken into account on the
basis of the measured distributions over the reactor of the spectral
characteristics, but in the practice of operating RBMK reactors the method
adopted has been direct periodic calibration of each radial monitor from the
power of the fuel assembly by scanning the assemblies with a hollow central
casing in an operating reactor, using beta emission detectors of the radial
monitor type or tri-axial fission chambers.
The theoretical-experimental dependence of the radial monitor's
sensitivity D to the neutron flux on the integral radial monitor current
Ii is an extremely efficient measure of the neutron fluence as a weak
function of the neutron spectrum and temperature of the monitor.
The ratio F,TD of the power of an assembly with monitor to the
neutron flux density at the site of the monitor is a function of the integral
energy yield of the assembly Ei.
When the reactor is in operation the power of the assembly with monitor
is calculated from the following equation:
Vs./Ls, K2f1: ? ai).fr
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where: Krpi is the individual calibration coefficient of the i-th radial
J monitor;
Ji is the current of the i-th radial monitor.
When the monitor is replaced in a spent fuel assembly, there is a
change of the present value of D in the "Skala" computer system with
variation in the integral radial monitor current in the computer memory.
Design calculations showed that the error associated with the use of Eq.(1)
when replacing a spent monitor by a fresh one was not more than 1%.
Generalized experience gained in operating RBMK reactors shows that the
Indicated calculation of the burnup of the radial monitor and assemblies does
not create errors of more than 1% in determining the power, density of Lhe
reactor.
The radial power density monitor (without the cabling) is installed in
the central casing of the fuel assembly with the aid of the central hall
crane. The cables are laid when the reactor is being assembled. Replacement
of failed monitors is carried out by remote control either with the reactor
shutdown or in operation after the monitors have been disconnected from their
cables. The laying of new cables is only possible with the reactor shut down.
. The radial power density monitor is designed to operate throughout the
service life of the fuel assembly. Experience with RBMK reactors has shown
the high reliability of these monitors. The mean time to failure, according
to operational data, is 9.7 x 104 h.
The radial power density monitor is considered to have failed in the
following instances:
- The emitter breaks off and as a result there is no current at the
? monitor connecting joint;
- The monitor readings in the "Skala" computer system are rejected
when making operative calculations with the "Prizma" program;
- There is a drop in the monitor's sensitivity, allowing for
burn-up, of more than 15% between two calibrations;
- There are rapid fluctuations in the monitor's signal that are not
confirmed by the neighbouring monitor readings;
The resistance of the monitor's insulation drops below 100 kohm.
Vertical power density monitors
To monitor the vertical power density distribution in the reactor use
Is made of 12 sets of monitors uniformly arranged in the core in the area of
? the radial distribution plateau. Each set contains seven beta emission
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detectors, with a silver emitter, arranged in a uniform vertical pattern, and
made like the radial monitors, in the form of a cable (type KNMS)(S). Each
sensitive element (section) of the vertical monitor is a spiral made from
cable with an outside diameter of 62 mm and height of 105 mm. The overall
length of cable in the spiral is 2.6 m. The centres of the top and bottom
sections are shifted 500 mm towards the centre with respect to the core
boundaries.
The design of the vertical monitor is shown in Fig. 2.42. Seven
sensitive elements are housed in a dry leaktight casing made of
corrosion-resistant steel, which is mounted in a channel similar to-the one
intended to hold the control rods. On the outside the casing is cooled by
stream of flowing water 7 mm thick, with a temperature at the reactor outlet
of not more than 70?C. A central tube running along the axis of the casing
is intended for periodic calibration of the sensitivity of the monitor
sections with the aid of a tri-axial fission chamber moving&vertically up the
monitor. 'In the non-working position the fission chamber may be left in the
central tube of the monitor since its sensitive volume will then lie below the
lower boundary of the core.
The sensitive elements are joined by high-temperture cables (type
KNMS (S)) to leaktight joints at the point whett the casing comes out in the
central hall. The same cable, enclosed in a protective sheath of corrosion-
resistant steel, is used to connect the sensitive elements via the joints to
an outside terminal block. The cable route is designed to ensure the best
immunity from interference. It is not,permitted, for example, to lay vertical
monitor cabling together with the cables feeding the control rod drives.
The inside of the casing is filled with an argon-helium mixture to
reduce radiation heat-up of the vertical monitors; the maximum temperature of
the sensitive elements does not therefore exceed 1500C.
To protect the space above the reactor from ionizing radiation coming
from the core and steam-water communication lines, the vertical power monitor
is fitted with two steel shielding plugs located at the top of the monitor
casing. Furthermore, the top of the monitor makes provision for the mounting
of a special protective cap, the function of which is to protect the monitor
joints from mechanical damage at the same time.
When the reactor is working at nominal power the currents from the
various sensitive sections of the vertical monitor may vary between a few
and 15 pA, depending on their position in the core.
The design of the set of monitors and the channel makes it possible to
replace them while the reactor is in operation as well as when it is shut
down. This is done by remote control using the crane of the central: hail.
The cables are laid while the reactor is being assembled, and replacement is
only possible with the reactor shut down.
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While the reactor is in operation, the signal from each section of the
vertical monitor permits calculation of the neutron flux density at iLs point
of location:
(n24)1) /(21is.'..p(Iii)".7841 (2)
where: n is the neutron density;
vo= 2200 m/s;
Kroj is the individul calibration coefficient of the i-th section of
the j-th vertical monitor;
1:,(1ii) is a correction for emitter burn-up, identical to the one
used in the radial monitor, and depending on the integral current of
the i-th section of the j-th vertical monitor Iii;
is the current of the i-th section of the j-th vertical,monitor.
The assumed lifetime of a vertical power monitor is 2.5 years.
Experience gained in operating RBMK reactors shows the satisfactory
reliability of the monitor. The mean time to failure derived from operational
data is 4.0 x 104 h.
The vertical power monitor is considered to have failed if there are
two defective sections side by side in it or if any three sections are
defective.
A section of the monitor is considered to have failed in the following
instances:
There is a drop in the sensitivity of the section, allowing for burnup,
of more than 15% between two calibrations;
There are rapid fluctuations in the section signal not confirmed by the
readings of neighbouring sections;
The resistance of the section insulation drops below 100 kohm.
Power density monitoring equipment
?Purpose and composition of equipment
The power density monitoring equipment is designed in the form of four
racks on which are placed the main functional units and systems of control and
monitoring. The equipment also consists of scanning instruments for operative
monitoring of the power density distribution in the reactor; a recorder and an
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indicating instrument for monitoring the thermal power; signal setting control
desk, switching devices and numerical indicators showing the co-ordinates of
the monitor called on for a signal to the scanning instruments, and rigs for
testing and adjusting the main functional units.
In terms of its purpose the power density monitoring equipment can be
divided into two parts: the equipment of the system for physical monitoring
of ihe radial power density distribution and the equipment for the system for
physical monitoring of the vertical power density. The principal functional
units in both parts are different in design: in the first case there are two
measuring devices, and in the second only -one such device. All three
measuring devices are served by one multi-channel recorder which records on a
numerical printer the values of the monitor output signals exceeding the
"emergency" level, the time of occurrence and the monitor's co-ordinates.
The measuring devices perform the main functions of shaping the
information signals and operatively monitoring the power density distribution
in Lhe reactor. These functions are carried out by separate circuits for each
radial and vertical monitor (monitoring lines). The measuring devices of the
system for physical monitoring of the radial power density process signals
received from 130 radial detectors mounted in the reactor, and there is a
possibility of switching in another 14 radial monitors (in all 144 monitoring
lines). The measuring device of the vertical power density monitoring system
processes signals from 12 seven-section vertical monitors (84 monitoring
lines), but has the possibility of processing signals from 12 eight-section
vertical monitors (in all 96 monitoring lines).
The radial and vertical monitor signals reach the inputs of the
individual amplifiers with adjustable negative feedback, which then transform
the monitor currents (input signals) into dc voltage signals (output
signals). These signals are fed to the inputs of the device for operative
monitoring of the power density distribution (signalling device), to the
inputs of the switching device (scanning instruments), to the unit averaging
the monitor signals, to the "Skala" computer system and, through the contacts
of the actuator relays, to the multi-channel recorder.
The signalling device compares the radial and vertical monitor output
signals with the set limit values of the signals. The comparison is made at
three levels (signal thresholds) termed the "undershoot", "warning" and
"emergency" levels. When the radial and vertical monitor output signals
deviate from the given levels, the relevant signal system is triggered in the
power density monitoring equipment and colour signals light up on the
CPS-power density monitoring system mimetic board. A green light means that
the radial and vertical monitor signal is equal to or less than the
"undershoot" level. Absence of a light means that the signal is greater than
the "undershoot" level, but less than the "warning" level, i.e. it'lies within
the accepted limits. A red light means that the radial or vertical monitor
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signali,is equal to or greater than the "warning" level, but has not yet
reached the emergency level, while a blinking red light means that Lhe signal
has attained or overshot the "emergency" level. In the latter case a sound
signal is, given along with the blinking light.
To present the information more clearly, the CPS-density monitoring
system mimetic board takes the form of a mnemonic representation of a
horizontal cross-section of the reactor in which there are indicators showing
the position of the control rods and the signal elements of the radial and
vertical monitors. The arrangement of the position indicators and signalling
elements on the mimetic board matches the arrangement of the control rods and
the radial and vertical monitors in a radial plane of the reactor. The "
systems for displaying the monitoring information enable the operator to sec
clearly the area in which there has been deviation of the monitor signal from
the set level, to determine from the type of signal how to actuate the control
rods (upwards or downwards) in order to eliminate the deviation, and to select
the rod, required for that purpose.
Furthermore, the radial equipment provides for the possibility of
altering, within +15%, the "undershoot" and "warning" levels for all the
radial monitors at the same time from the operator's control desk; this
enables operators to detect promptly of the areas where the power density is
close to maximum.
The radial monitoring equipment provides for two modes of operation:
comparison of the radial monitor output signals with the floating levels
(thresholds) of the "undershoot" and "warning" signals and with the fixed
levels (thresholds) of the "emergency" signal; and a mode by which the radial
monitor output signals are compared with the fixed thresholds for all three
levels.
In the first mode of operation the "undershoot" and "warning" signal
levels for each radial monitor vary in proportion to the arithmetic mean of
the radial monitor ouput signals, i.e. in proportion to the present power of
the reactor, while the "emergency " level is fixed at a level selected on the
basis of operational requirements. When the arithmetic mean of the radial
output signals attains a pre-set maximum (set level of power at the given
stage of reactor operation), the "undershoot" and "warning" signal thresholds
are fixed (limited) and the radial power density monitoring equipment
automatically switches to the second operational mode.
The vertical monitoring equipment operates in a mode where it compares
the output signals of the vertical monitor sections with the floating
thresholds for all three signal levels, i.e. it only carries out relative
monitoring of the vertical power density of the reactor. Here, the signal
thresholds for each vertical monitor vary in proportion to the arithmetic mean
of the signals from the sections of the given vertical monitor.
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To check the working capacity of the radial and vertical ii'lonitors the
equipment makes provision for devices by which to determine the resistance of
the insulation of any detector by switching an additional resistor Rgob
(100 kohm) into the input circuit of an individual amplifier di' the monitoring
line. From the relative decrease in monitor output signal the resistance of
the monitor insulation can be computed:
g
u7i,
(3)
where: v is the monitor output signal up to the moment when the Rgob is
switched in;
v' is the monitor output signal after the Rgob has been switched in.
Principal specifications of the equipment
The
amplifiers
negative.
individual
resistance
maximum value of the signals at the output of the individual
of the monitoring line is 5 V, while the polarity of the signals is
The range of adjustment of the coefficients for transformation of
amplifiers ranges from 0.26 to 0.78 V/pA. The input amplifier
is not more than 100 ohm.
The principal relative error transforming the input signals (in
percentage) is not more than:
4 5. om .730.ce ' ill)
where: Jmax = 19 pA is the maximum input current;
J is the present value of the input current (pA).
(4)
The permissible capacitance of the monitor together with the cabling
should not exceed 0.05 pF.
The permissible resistance of the load on the output terminals of the
amplifiers is at least 2 kohm.
The equipment ensures that eight monitor signals are displayed
simultaneously on indicating instruments (141830A): one of the radial monitor
output signals and seven output signals from the sections of the vertical
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monitor selected. Calling for signals on the indicating instruments is
effected by switching devices which store the radial and vertical monitor
address in a code for the co-ordinate grid of the reactor channels.
The equipment shapes a signal equal to the arithmetic mean of the
radial monitor output signals (reactor power signal) and displays it on an
indicating instrument (0-100 pA scale) and on a recording instrument
(0-100 my scale; scale run-through time 10 sec). Provision is made for the
iro_roduction of a correction to the arithmetic mean of the signals that takes
into account the absence of signals at the amplifier input. The mean relative
error in shaping the averaged signal, given signals of at least 2.5 V at the
individual amplifier output and a number of average signals ranging from 70 to
130, does not exceed 40.5%
The equipment shapes four power signals for the quadrants of the
reactor equal to the arithmetic mean of the radial monitor output signals of
the corresponding quadrant; these are displayed on four indicating instruments.
The equipment shapes for each vertical monitor a signal equal to the
arithmetic 11. Lht cutput. sisnalc, c..;7 :! D
d v)
(19)
where x is a normal distribution quantile corresponding to a given
probability of non-acceptance of the true measurement;
0
v .
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The readings of these monitors are not used in the calculations; ,information
about them is automatically fed to the printer for use by the operating staff.
Finally, the power of each assembly, including the assemblies with
monitors, are calculated from the equation:
V(tc,.501)7Ifei
(21)
Summing is conducted over the four monitors closest to the i-th ass'embly. The
weight coefficients bik are, determined by solving a set of four linear
equations compiled with the requirement of, minim= 'error. These coefficients
depend on the distance between the assembly and monitor, and the statistical
characteristicvalues.ofthe error in calibrating the monitors.
vj
As well as the calculation of the power-Wi of each assembly, there is
also calculation of the dispersion Di of the error in it.
Calculation of the power margin coefficient
The maximum permissible power of the RBMK reactor fuel assembly is
taken to be the power at which the probability of the assembly experiencing
critical heat flow attains a preset value constant in time and identical for
all assemblies. In accordance with this definition the power margin up to
critical heat flow for the i-th assembly is:
(22)
where: Wi is the power of the fuel assembly;
Wkri is the power of the assembly at Which there is onset of critical
heat flow. This value is determined from the tables as a function of
the water flow through the channel, pressure in the drum separator and
wilier temperature in the pressure header;
- 4.1)1-n
(23)
e21-= 7J7/73 ;
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where: x is the normal distribution quantile corresponding to the set
probability that the assembly will experience critical heat flow;
Di is the dispersion of the relative error determining the assembly
power;
DTni is the dispersion of the relative error in determining the
critical power of the assembly (methodological);
DTn2 is the dispersion of the error in determining the reactor power;
DTn3 is the dispersion of the determination of Wkri through errors
in measuring the flow, pressure and temperature.
Calculation of the graphite temperature
The temperature of the graphite is calculated on the basis of assembly
power calculations given for the vertical power distribution and signals from
thermocouples mounted in the stack.
The graphite temperature is calculated for each k-th join in the graphite
columns from the following relationship:
z e oe vv,? yor
r K
where tT is the mean temperature of the coolant in the reactor;
14/
Wic =
rr/ .
Of) Pr?
(v) 3
t= I
(24)
(25)
where: Wi(k) is the assembly power in the channel adjoining the k-th join;
in is the number of these assemblies (m 4);
yi(k) is a coefficient taking into account the effect on heat
removal from the graphite by channels with different loadings (with
fuel, absorbers, and control rods);
wk is the relative neutron flux density at the level where the
thermocouples are located;
is a coefficient of proportionality between the thermocouple signal
and the power as determined by the least square method.
The thermocouple readings which stand out strongly are processed.
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Calculation of the in-core monitor settings
The setting for the radial monitor is calculated from the equation:
7 mayI.
J ? ?# W.
e
(26)
where: Jj is the signal from the j-th monitor;
Wi is the power of the i-th assembly in the region Wj close to the
j-Lh monitor (this region is a square 5 x 5 reactor cells in size, in
the centre of which the j-th monitor is positioned);
q(ri) is the relative regulation value of the power of the i-th
assembly;
C is the normalizing constant corresponding to the prescribed maximum
assembly power.
The setting for the vertical monitor is calculated from the requirement
that the prescribed maximum linear load on the fuel element and the assemblies
close to the monitor should not be exceeded.
From the design values of the settings for the in-core monitors are
calculated and printed out, on request, the transmission coefficients of the
amplifiers of the comparator units in the physical power density monitoring
system, which are then fed into the equipment.
Calculation of the operative reactivity reserve
The operative reactivity reserve on the control rods is calculated from
the equation:
hie
A Pe
rtpy,
eie sb.,z(ori 2 see ja,_
t 0
(27)
where: Ck is the relative "weight" of the rod and depends on its type;
Npc is the number of regulating rods;
.k(z) is a value proportional to the vertical distribution of the
absolute neutron density flux at the point where the k-th rod is
situated, and is calculated by the use of the values of the assembly
power Wi computed from the vertical monitor readings averaged over
the reactor.
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2.9.9. Presentation of information on the calculation results
All the calculation results are transmitted, upon call by the staff, to
printing devices in the form of charts (cartograms) and summary tables. The
latter show, in particular, lists of 60 channels in which the channel power is
maximal, 60 channels with maximum graphite temperatures, and 60 channels with
the least margin coefficients.
Transmitted automatically to the printout device are the time, the
co-ordinates of the rejected detector, the rejection constant, the time, the
channel co-ordinate and the power (in the case where the prescribed value is
overshot).
The mimetic board for the channels shows the channels in which the
power is greater than the value set by the operator; the channels in which the
coefficient is below the operator's setting, and so forth.
The values of any quantity computed can be shown on the digital
Indicating instruments if called for.
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Fig. 2.36- Thermocouple unit and assembly: _
1. Tube; 2. Rod; 3. Graphite bush;
4. Chromel?Alumel heat sensor.
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c?i
II
Rse.
/04
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OR
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. "From.gas scrubbing
system
? .
6
Bleed
if
Ventilation
dF z
Fig.
To group.
valve
Ne(
(a) Process channel integrity monitoring system;
(b) Chromel?Copel temperature sensor unit.
1. Reactor; 2. Process channel; 3. Impulse tubes; 4. Panel; 5. Temperature
sensor; 6. Retrieval device; 7. Digital display equipment; 8. "Skala" system;
9. Channel mimic hoard; 10. Humidity hoard; 11. Humidity indicator;
12. Group valve; 13. Humidity sensor; 14. Gas blower.
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tr)
/70
014
A
06
A-A
L'151
Fig.2.4I
Design of the radial power density monitor
Iju
06
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0 245
0 75 .
2
0 68
I2
Fig. 2.42 Design of the vertical power
density monitor: 4) cable;
(2) leaktight tube;
(3) sensitive elements
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2.10. ":"Safety systems
?
2.10.1. Protective safety systems
2.10.1.1. Reactor emergency cooling system
The emergency core cooling system (ECCS), shown in diagram 2.43, is a
protective safety system designed to draw off the residual heat release (after
suppression of the chain reaction) through the timely feeding of the
appropriate volume of water into the reactor channels in the event of
accidents accompanied by damage to the core cooling system.
Associated with such accidents are: ruptures in the large diameter
multiple forced circulation circuit (primary coolant circuit) ? (MFCC)
pipelines, ruptures in the fresh steam pipelines and ruptures in the feedwater
pipelines.
In addition to this, the ECCS may be used for the emergency feeding
of water into the reactor channels in situations which are not connected with
ruptured pipelines, but which make it impossible to supply the water through
regular systems (for example, steam in the electric feed pumps).
The ECCS was designed with the following requirements in mind:
1. It must ensure a water supply to the damaged and undamaged
halves of the reactor in quantities not less than those shown in
diagram 2.44, thereby preventing melting, massive overheating
and seal failure in the fuel elements;
2. The ECCS must come into operation automatically on receipt of
the "maximum design-basis accident signal", which must
distinguish between the damaged and undamaged halves and be
formed on the basis of the following indications:
(a) An increase in pressure in compartments containing MFCC
pipelines (indication of pipeline rupture);
(b) Coincidence with either of the following two *signals
(showing selection of the damaged half):
- Drop in level in the steam separators of the damaged
half of the reactor;
- Decrease in the pressure differential between the main
circulation pump pressure header and the steam
separators of the damaged half of the reactor;
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3. The speed of operation of the ECCS is such that the break in
water supply to the damaged half of the reactor in the event of
a maximum design-basis accident does not exceed 3.5 sec.;
4. There must not be an unacceptable reduction in waterY supply to
the reactor channels as a result of its unproductive loss
through the point of rupture in the compartment;
5. The system must perform its safety functions in the event of any
failure, occurring independently of the source event,, in any of
the following parts of the system: any active or passive
element having moving mechanical parts;
6. The system must comprise a number of independent channels
(subsystems) and must function with the required effectiveness
in the event of a failure, occurring independently of the source
event, in any one channel (subsystem) of this system;
7. In the event of drainage of the ECCS vessels, nitrogen from them
must not be allowed to reach the reactor;
8. The ECCS must operate as intended in the event of a maximum
design-basis accident coinciding with a loss of own-requirement
power from the power unit.
In order to comply with the above essential requirements, the ECCS
comprises three independent channels (subsystems), each of which ensures not
less than 50% of the required output.
Each channel (subsystem) includes a fast-acting part and a part
providing prolonged afterheat removal.
The fast-acting part supplies the required quantity of water to the
channels of the damaged half of the reactor during the initial stage of the
accident
The fast-acting parts of two ECCS channels (tile cylinder parts) take
the form of a system of vessels (filled with water and nitrogen at a pressure
of 10.0 MPa), connected by pipelines and headers to the distributing group
header of the primary coolant system.
A Du 400 gate valve is used as a quick-closing armature for bringing
into operation the cylinder parts of the ECCS; by this means the required
supply can be brought to the damaged half of the reactor in the space of
3.5 sec. The power supply for the gate valves is supplied under the maximum
reliability category by the accumulator batteries (see section 2.7.3).
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Each of the two cylinder parts comprises six vessels of 25 m3
volume. The total initial volume of water of one cylinder part is
approximately 80 m3, and of nitrogen, approximately 70 m3. Each cylinder
part supplies not less than 507. of the required quantity of water to the
damaged half of the reactor over a period of not less than 100 sec. The
period of operation depends on the scale of the coolant leak from the primary
coolant system in the event of an accident.
The configuration of the cylinders is symmetrical, in order to reduce
the "collector effect" when discharging.
The nitrogen from the ECCS vessels is prevented from_reaching the
reactor through the automatic closing of two gate valves installed in series
on the pipelines from the cylinder parts to the distributing group header on
receipt of a signal indicating the minimum level in the cylinders.
The fast-acting part of the third ECCS channel is a unit for
supplying water from the electric feed pump, which ensures a supply of not
less than 50% of the required amount of water to the damaged half of the
reactor. In the event of a maximum design-basis accident coinciding with a
loss of power to users of the own-requirement supply from the power unit, the
supply of water from the electric feed pump is assured for a period of
45-50 sec. as the pumprruns down in tandem with the turbogenerator.
The reserve power for the drive units of the fast-acting gate valves
is supplied by the independent sources of uninterrupted power (the accumulator
batteries).
The prolonged afterheat removal part provides cooling both to the
damaged and undamaged parts of the reactor. It comes into operation no later
than the moment at which the fast-acting part of the ECCS ceases to operate.
The long-term afterheat removal part of each of the three ECCS
channels comprises two groups of pumps:
the cooling pumps of the damaged half of the reactor;
the cooling pumps of the undamaged half of the reactor.
The pump section of the cooling system for the damaged half of the
reactor of each of the three ECCS channels consists of two pumps connected in
parallel. Its function is to ensure a supply of water at a rate of
approximately 500 t/h, that is, not less than 50% of the required rate for the
damaged half in the event of a maximum design-basis accident.
The water is drawn by the pumps from the pressure suppression pool of
the accident localization system, is cooled by the service water in the heat
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exchanger mounted on the common intake line of the two pumps, and reaches the
ECCS headers through the discharge lines.
Flow-limiting inserts are mounted on the discharge lines of the
pumps; they are designed to ensure the steady functioning of the pumps in
emergency situations characterized by a sharp drop in pressure in the
reactor's coolant circuit through a ruptured pipe (flow limitation is achieved
by boiling water in the narrow cross-section of the insert).
The pump section of ?the cooling system for the undamaged half of the
reactor of each of the ECCS channels contains one pump and supplies water at a
rate of approximately 250 t/h, that is, not less than 50% of the flow required
for the undamaged half in a maximum design-basis accident.
Water is drawn by suction from the tanks containing clean condensate
and, by means of the discharge line, reaches the headers of the cylinder
section behind the quick-closing armature.
The flow-limiting inserts in the discharge lines of the pumps perform
the same functions as for the pumps of the cooling system of the damaged half
of the reactor.
Stand-by power for the electric motors of the pumps and armature
drive units of the pump sections of the damaged and undamaged parts of the
reactor is supplied by the diesel generators.
2.10.1.2. System to protect against excess pressure in the main coolant
circuit
This system is designed to ensure that the permissible pressure level
in the circuit is not exceeded; it does this by drawing off the unbalanced
steam into the pressure suppression pool, where it is completely condensed.
The system includes a pulsed safety device and a system of pipes and headers
which conduct the steam into the pressure suppression pool of the accident
localization system. The pulsed safety device comprises the pulse valves and
the main safety valves.
The system satisfies the following main requirements:
- It ensures that the pressure in the circuit is not exceeded by
more than 15% of the working pressure, taking into account a
single failure in the system of an active or passive element
with moving mechanical parts;
- It comes very reliably into operation when the pressure in the
coolant circuit reaches the minimum operating value;
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It is very reliable in closing the main safety valves when the
close value is reached;
It is capable of working adequately under conditions of
alternating dynamic loads upon operation of the main safety
valves;
It introduces steam into the water of the pressure suppression
pool at speeds that are close to that of sound, even when one
main safety valve is in operation (this is necessary for
shock-free steam condensation).
, A schematic drawing of the system for discharging steam from the main
safety valves into the pressure suppression pool of the accident localization
system is shown in diagram 2.45.
The system comprises eight main safety valves with a total output of
5800 t/h, under nominal circuit pressure, i.e. an output which is equal to the
nominal steam output of the reactor installation. The control of each main
safety. valve with an output of 725 t/h is effected using a directly acting
pulse valve (lever-gravity type), equipped with an electromagnetic drive unit
for opening and closing.
Steam from the main safety valves is discharged into the pressure
suppression pool beneath the water level through submersible nozzles, each
with an exit diameter of 40 mm (approximately 1200 nozzles in all).
In order to prevent the formation of a vacuum in the discharge
pipelines and the consequent entry of water into them, and also to ensure the
shock-free condensation of possible small flows of steam through the closed
main safety valves, steam-air ejectors are used.
The steam discharge systems include:
Control of the absence of water level in the pressure
suppression pool headers;
Control of the temperature conditions of the external surface of
the pipes for each main safety valve and pipes situated inside
the pressure suppression pool.
When the unit is working normally, the main safety valves are closed
an& the system is in waiting mode.
The system comes automatically into operation only when there is
excess pressure in the primary coolant circuit; it does this as follows:
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76 kgf/cml - 1 main safety valve operates;
77 kgf/cm2 - 2 main safety valves operate;
78 kgf/cm2 -1 main safety valve operates;
81 kgf/cm2 - 4 main safety valves operate.
It is possible for the operative staff of the unit control room and
reactor control room to forcibly open the main safety valves.
The main elements of the system to protect the circuit from excess
pressure during operation have undergone experimental bench tests-.,. The system
as a whole underwent comprehensive and direct testing in respect of design
requirements during the period of commissioning operations.
r
, 2.10.1.3. System to protect the reactor space from excess pressure
,
,
The purpose of this system is to ensure that the permissible pressure
in the reactor space is not exceeded in an accident situation involving the
) rupture of one fuel channel. It achieves this by drawing the steam and gas
4
1 mixture from the reactor space into the steam and gas discharge compartment of
the pressure suppression pool and subsequently into the pressure suppression
pool itself.
The system satisfies the following main requirements:
I
- It ensures that the excess pressure in the reactor space does
not exceed 1.8 kgf/cm2 (abs.) in the event of a total cross
break of one fuel channel and taking into account a single
failure in the system of a passive element with moving
mechanical parts (there are no active elements in the system);
- It prevents water from the steam and gas discharge compartment
of the pressure suppression pool from entering (overflowing
into) the reactor space in the event of a maximum design-basis
accident;
- It ensures that the reactor space is reliably isolated from the
atmosphere.
A schematic drawing of the system for protecting the reactor space
from excess pressure is shown in Fig. 2.46.
The reactor space is constantly connected to the steam and gas
discharge compartment of the pressure suppression pool of the accident
localization system by eight Du 300 pipes (four pipes below and four pipes
above the reactor space, which then join to become two Du 600 pipes).
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Each of the Du 600 pipes goes to its own level of the compartment and
is immersed in 2 m of water, that is, under normal unit operating conditions,
the reactor space is separated from the atmosphere by a hydroseal 2 m high.
The height of the vertical sections of the Du 600 steam discharge
pipes from the reactor to the water level in the compartment is more than
28 m, and for this the Du 600 pipe, which joins together the four Du 300 pipes
beneath the reactor space, rises specially to the 28.8 mark and then
immediately drops into .the compartment.
, Such an arrangement is necessary in order to prevent water or a
steam-Lair mixture from the compartment from entering the reactor space in the
event Of accidents involving the rupture of MFCC pipes, right up to a mliximum
design-basis accident.
The volume of water in the compartment is selected and maintained in
such a way as to ensure a sufficient reserve to fill the steam discharge pipes
in the situation indicated above.
A second and supplementary barrier, designed to prevent water or
steam-air mixture from the compartment from entering the reactor space, is in
the form of non-return (escape) valves, which make it possible to discharge
steam from the compartment into the pressure suppression pool and to prevent
its reverse flow.
In order to prevent the [un-?) controlled spread of solid radioactive
wastes throughout the water content of the pressure suppression pool, the
steam-gas discharge compartment is reliably cut off (by three barriers) from
the water content of the pressure suppression pool in the event of a fuel
channel rupture.
In the event of a rise in pressure in the reactor space to
1.2 kgf/cm2 (abs.), the hydroseal in the compartment pops open and the steam
and gas mixture enters the compartment through the steam discharge pipes.
Should the pressure in the above-water part of the compartment reach
1.1 kgf/cm2 (abs.), the non-return (release) valves open and the steam and
gas mixture enters the steam distribution corridor; it then enters the water
of the pressure suppression pool by means of the steam discharge pipes. The
steam formed in the reactor space in the event of a fuel channel rupture is
fully condensed, initially in the water in the compartment, and, when the
storage capability of this is exhausted, in the pressure suppression pool.
The gas from the reactor space, bubbling through the layer of water in the
compartment/pressure suppression pool, is cooled and maintained in the
compartments of the accident localization zone, after which, following the
necessary period of holding and cleaning, it is discharged into the atmosphere
by the hydrogen disposal system.
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The maximum pressure in the reactor space at all stages of the
accident sequence does not exceed 1.8 kgf/cm2 (abs.).
The protection system includes:
Pressure control (underpressure) in the reactor space;
- Control of the level of steam-gas discharges in the compartment;
- Reliable drainage of the steam discharge pipes.
Monitoring of the process parameters and control of the active
elements of the system (cut-off armature) is carried out by operative staff in
the unit control-room and reactor control-room.
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2.10.2. Safety based on confinement systems
The accident confinement system built in the fourth unit is designed
? to confine radioactive releases during accidents involving failure of any
piping of the reactor cooling circuit, except the steam-water lines, upper
fuel channels and the part of the downcomers which is located in the drum
separator area. The schematic diagram of the system is given in Fig. 2.47.
2.10.2.1. System of sealed locations
The basic component of the confinement system is the system of
sealed, locations comprising the following locations of the reactor part:
Strong leak-tight compartments (1 and 2 in Fig. 2.47) arranged
symetrically in relation to the reactor axis and designed for
an overpressure of 0.45 MPa;
Locations of the distributing group header and .lower water
communication lines (DGH-LWCL) (3 and 4), which are also
symetrical in relation to the reactor axis and are separated
from each other by the reactor's supporting cross-piece having
a total non-tight area of 5 m2. According to the strength
specifications for the reactor structural elements, these
locations do not tolerate a pressure rise above 0.08 MPa and
are designed for this value. The strong leak-tight
compartments and the DGH-LWCL locations contain all those
reactor circuit elements that may be damaged in accidents for
which ?the system is designed;
- Location of the steam distribution corridor (5);
- Location of the two-storey condensation-type pressure
? suppression system, a part of which is filled with water (7)
and the rest with air (8).
The sealed locations are connected with each other by means of
valves of three types:
Non-return valves (9), installed in the openings of the cover
separating the DGH-LWCL location and the steam distribution
corridor;
Release valves (10), installed in the openings of the cover
separating the air space in the pressurizer relief tank and the
strong leak-tight compartments;
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Panels of non-return valves (11) installed in the partitions
separating the steam distribution corridor and the strong
leak-tight compartments.
The locations of the strong leak-tight compartments , .and steam
distribution corridor are connected with the water volume of the
condensation-type pressure suppression system by steam outlet channels?(17).
In normal operation the system of sealed locations and the
condensation-type pressure suppression system .operate in the stand-by mode.
In emergency situations the system functions in the following
manner. If a reactor circuit component in one of ?the strong leak-tight
compartments fails, the boiling coolant begins to flow into that
compartment. Steam formation leads to a rise in pressure in the accident
location. ,The non-return valves of the panels connecting the damaged half of
the compartment with the steam distribution corridor (11) open at a pressure
difference of > 0.002 MPa. When the pressure in the damaged half of the
compartment attains a value sufficient for displacing the liquid column from
the steam outlet channels, the steam and air mixture begins to flow into both
stories of the condensation system at the same time. By bubbling through the
water layer the steam condenses and the air is collected in the air space of
the condensation system; when the pressure there reaches > 5 kPa, the release
valves connecting the air space of the condensation system with the undamaged
strong leak-tight compartment open and part of the air flows to that
compartment. Thus, its volume is used in this emergency situation to reduce
pressure in the damaged half of the leak-tight compartment. In the course of
such an emergency the non-return valves (9) remain closed.
If the reactor circuit failure occurs in the DCH-LWCL location, the
pressure rise there opens the non-return valves connecting it and the steam
distribution corridor (at a pressure difference of > 0.02 MPa). From the
corridor via the steam discharge channels the steam-air mixture goes into the
water volume of the condensation system's central part situated under the
corridor. Pressure rise in the air space of the condensation system'opens
the release valves connecting that space with the two strong leak-tight
compartments. In this kind of emergency situation the volumes of both
compartments are used to reduce pressure in the damaged location, while the
panel valves (11) remain closed.
All the sealed locations of the system, except the condensation-type
pressure suppression system, have a 4 mm-thick lining of the VST3KP2 steel
and are subjected to control tests for local and integral leak-tightness.
The condensation-type pressure suppression system has a 4 mm lining of the
08Kh18N1OT [crl8nil0ti] steel. Figures 2.48 and 2.49, respectively, show the
results of calculation of pressure changes in the sealed locations during an
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accident involving a rupture of the main circulation pump pressure header
(average diameter 900 mm) in the strong leak-tight compartment and those
during an accident with rupture of the distributing group header (average
diameter. 300 mm) ?in the DCH-LWCL location. As will be seen from the graphs
in these figures, the overpressure in the damaged leak-tight compartment does
not exceed the maximum permissible value of 0.25 MPa, while the overpressure
in the -?damaged DGH-LWCL location does not exceed the maximum permissible
value of 0.08 MPa.
The system carries out its functions under conditions of a single
failure of any passive component having moving parts (the system has no
active components).
2.10.2.2. Penetrations and doors
?
To prevent the spread of activity outside the sealed locations the
sealed barriers of the accident confinement system (walls, floors and
ceilings) are equipped with special sealed penetrations at the places where
they are traversed by pipes or electrical cable.
The pipe penetrations are designed to withstand the action of jets
from a pipeline during its complete rupture. In such a case, the sealing of
the penetration is not destroyed.
The design of the penetrations is such as to allow checking of their
tightness during both assembly and operation. The penetrations ensure
leak-tightness at an overpressure of up to 45 kPa in the accident confinement
locations, at a temperature of up to 1500C.
The sealed pipe penetrations intended for the passage of "hot" lines
are equipped with a water or air cooling system to prevent overheating of the
concrete in the penetration area.
Sealed doors are intended to provide access of the service personnel
into the locations of the accident confinement system when the reactor is
, shut down and to ensure sealing of those locations when the reactor is
operating.
The sealed doors of the condensation-type pressure suppression
system ensure the necessary leak-tightness and operability after elimination
of each accident situation, including a design basis accident (DBA).
There are two sluice-gate-type entrances into the above pressure
suppression system, each entrance having two successive sealed doors.
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2.10.2.3. Cut-off and sealing valves
The cut-off and sealing valve system ensures isolation of the
accident confinement area by cutting off the communication lines between the
sealed and non-sealed locations.
The system's design is based on the following major principles:
- All communication lines traversing the sealing circuit which
have to be shut off at the time of an accident in order to
prevent escape of radioactive substances outside the sealed
locations are equipped with three successive cut-off devices;
- Each pipeline which is not connected directly with the primary
circuit or with the space of the sealed locations is equipped
with one cut-off device mounted outside the sealing circuit;
The positions of the shut-off valves which seal the locations
under accident conditions are indicated on the unit's control
board (safety panels) and stand-by control board, from where
they can be remote-controlled by the operator if necessary;
The drives of the cut-off valve installed on one main line are
powered from independent sources of the reliable power supply
system of category 1A.
?The special fast-acting (10-15 s) cut-off valves and non-return
valves are used for isolating and sealing the accident confinement locations.
Pre-operational tests are carried out in the manufacturing factories.
The testing of isolation valves during the operation of the nuclear
power plant is performed only when the unit is shut down. The entire system
of isolation valves is verified. The tests include verification of their
efficiency and tightness.
The valves are closed automatically by DBA signals.
The system of cut-off and sealing valves is so designed that any
single failure in the system will not disrupt its functions.
2.10.2.4. Condensation-type pressure suppression system
The purpose of this system is to condense steam formed:
- during an accident involving reactor circuit failure;
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during the actuation of the main safety valves;
during leaks through the main safety valves under normal
operating conditions.
The system is a two-storey reinforced concrete tank with a metal
lining' inside. The space in each storey is divided by longitudinal
partitions into four corridors and by transverse partitions into three
sections: two lateral (under the leak-tight compartments) and one central
(under the steam distribution corridor). The longitudinal and transverse
walls 'have the necessary openings for water and air. The lower part of each
storey is filled with water. The thickness of the water layer in each storey
is 1200 mm. The total volume of water in the two storeys is 3200 m3 and
the volume of the air space is 3700 m3.
Steam goes into the water volume through the steam discharge
? channels, which are located uniformly over the whole area of the leak-tight
compartments and the steam distribution corridor. Each steam discharge
channel is in the form of a pipe-within-pipe-type block, which ensures
simultaneous and uniform delivery of steam to both the storeys. The number,
diamete'r- and spacing of the steam distribution pipes and their depth under
water are determined from tests On a large-scale model and ensure full
condensation of the steam in the water volume of the condensation system, its
uniform 'heating and rapid reduction of temperature in the damaged sealed
location during accidents involving reactor circuit failures.
? The upper storey of the system has the necessary number of special
vertical overflow pipes with a diameter of 800 mm (28 in Fig. 2.47). The
purpose of these pipes is to maintain the necessary level in the upper storey
and to equalize pressure in the air spaces of both storeys.
There is continuous monitoring of the water level in both storeys
and of the temperature and chemical composition of water. The required
chemical composition of the water is ensured by the by-pass purification unit.
The confinement system also includes a system of heat removal from
the condensation system and from sealed locations and a system of hydrogen
removal.
Heat removal from the sealed locations of the confinement system is
carried out by two systems:
1. ? Sprinkler cooling system;
2. Surface-type condensers located in the steam distribution
corridor.
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The sprinkler cooling system carries out the following functions:
- Cooling and purification of for air in the leaktight
compartments and in the air space of the condensation system
under both normal operating and accident conditions;
- Cooling of the water volume in the condensation system.
The main components of the sprinkler cooling system are shown in
Fig. 2.47. Water is collected from the condensation system and is sent via
three lines of pipe (each accounting for 50% of the system's throughput) to
the heat-exchangers (15), where it is cooled by industrial water, and then
through pumps (14) to all users of the system:
- to the jet coolers (12) in the leaktight compartments;
- to the nozzles (13) located in the air space in the twp storeys
of the condensation system.
The jet coolers form part of the sprinkler cooling system and ensure
circulation of air in the leaktight compartments, cooling of the air and
removal of radioactive aerosols and steam from it.
The air from the upper (hottest) part of the leaktight compartments
is collected, cooled by water jets and sent to the lower part of the
compartments. After its contact with the air the cooling water returns to
the condensation system. The jet coolers work continuously both under normal
operating and accident conditions.
The sprinkler nozzles located in the air space of the condensation
system ensure spraying of the cooling water and mixing and cooling of the
air. The necessary pressure difference at the nozzles is created by the
reducer discs at the feeders supplying cooling water. The nozzles work
continuously both under normal operating and accident conditions.
The purpose of the surface-type condensers (16) in the steam
distribution corrider is to remove heat from the sealed locations during
accidents involving reactor circuit failure by condensing the part of steam
entering to the corridor. The cooling medium is industrial water. In normal
operation the surface-type condensors work in the standby mode and go into
operation on receiving the DBA signal.
During the development period the efficiency of the system was
confirmed by tests on a large-scale model.
The system performs its functions when there is a single failure of
any active component or passive component with moving parts.
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2.10.2.5. Hydrogen removal system
The purpose of this system is to create a negative pressure in the
accident confinement locations, to measure the concentration of hydrogen,
which may enter these locations with uncontrolled leaks from the multiple
forced circulation loop and also during steam discharge from the main safety
valves and during accidents associated with pipe failures in the main forced
circulation loop, and to remove the hydrogen upon its occurrence.
Under normal operating conditions of the unit, hydrogen may enter
the accident confinement system locations with coolant leaks, the magnitude
of which is taken as 2 t/h, and with possible leaks of steam through the
closed safety valves.
Hydrogen may also enter under conditions of short-time steam release
during actuation of the main safetey valves and under conditions of pipe
failure.
The highest quantity of hydrogen may enter the location under DBA
condifions (the hydrogen accumulated in the coolant and also that formed
during-the accident by radiolysis and by the reaction of zirconium with
water).
Figure 2.50 gives the total amount of hydrogen entering under these
conditions.
Against the existing standard of 4% (by volume) for the lower
explosive limit of hydrogen in air, 0.2% concentration (by volume) was taken
as the design value in the project. It is necessary to evacuate air from the
accident confinement locations at the rate of 800 m3/h in order to maintain
this concentration under the most unfavourable conditions. This rate is
adopted also for all other operating conditions of the unit. The hydrogen
removal system (Fig. 2.51) comprises: an electric heater, a contactor, a
condenser, a moisture separator and a gas blower. This equipment is divided
into three sub-systems, each of which is located in a separate compartment.
Each sub-system is provided with a stand-by active component-gas blower. The
protective isolation valves are located in a separate compartment. The
principle of 3 x 100% redundancy is envisaged.
Under normal operating conditions the gas-air mixture passes through
the electric heater, contactor (in the presence of hydrogen), condenser and
moisture separator and, by means of the gas blower, through the filtration
plant and is discharged into the atmosphere.
By a DBA signal the protective isolation valve closes and the-
equipment of the hydrogen removal system is disconnected. After 2-3 hours
(as hydrogen accumulates) the operator opens the protective isolation valve
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and switches on the gas blower of the hydrogen removal system. The control
'is exercised from the latter system's control board. The mode of
post-accident operation of this system is identical with operation under
normal conditions except that the mixture is discharged through the gas
activity reduction system (GARS). In addition, there is provision for
recirculation of the mixture.
If necessary (according to gas analyser readings), ini.ensified
evacuation from any location of the accident confinement system can be
organized by switchibng on the stand-by gas blower or the stand-by sub-system.
Nitrogen supply is provided for cleaning the hydrogen removal system
equipment and for fire fighting.
Gas analysers carry out continuous automatic monitoring of hydrogen
concentration in all locations of the accident confinement system. The
control board of the hydrogen removal system and the unit's control board
(operational part) have warning signals (acoustic and luminous) for rises in
hydrogen concentration in the accident confinement system locations. There
is also provision for control measurements of hydrogen concentration in these
locations by manual sampling on chromatographs. In the hydrogen removal
system proper the flow rates, temperature and radioactivity are measured.
All the data are displayed on the system's control board.
The monitoring and control system has three independent channels.
The devices of the hydrogen removal system receive their power supply from
the sources of the corresponding safety sub-systems.
The sprinkler system supplies cooling water to the condensers of the
hydrogen removal system.
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2.10.3. Power-supply safety systems
2.10.3.1. Power supply for the plant's own requirements (house power supply)
Own-requirement users at the plant (house users) can be divided up into
the following groups depending on the extent to which they need a reliable
power supply:
The first group comprises users which cannot tolerate an
interruption in power supply or can tolerate interruptions of
between fractions of a second and several seconds in any regime -
including the regime of total loss of alternating current from
working and reserve own-requirement transformers - and for which a
power supply is absolutely essential after a scram (tripping of
the emergency protection system);
The second group comprises users which can tolerate interruptions
in the power supply of between tenths of a second and tenths of a
minute in the same regimes and for which a power supply is
absolutely essential after a scram;
The third group comprises users which do not require a power
supply in the event of total loss of power from working and
reserve own-requirement transformers and which during normal
reactor operating conditions can tolerate an interruption in power
supply for the time needed for switching from the working
own-requirement transformer to the reserve transformer.
For own-requirement users at the plant there are two independent and
interchangeable (mutually redundant) power supplies: a normal working supply
and a reserve supply from the working and reserve own-requirement transformers.
For users of the first and second groups there is an additional supply
from a third independent emergency source.
The emergency power sources consist of the following:
(a) A storage battery with static transformers for users of the
first group;
(b) Automatic diesel generators for users of the second group.
The circuit diagram of own requirements is shown in Fig. 2.52.
2.10.3.2. Circuit diagram of own requirements of 6 kV for users of the third
group
The third group of users comprises the main circulation pumps, the feed
pumps, the first- and second-stage condensate pumps, the mechanisms of reactor
auxiliary-systems and of the machine room and other systems involved in
reactor operation in normal condition.
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For providing third-group users with power, the plant has 6 kV and
380/220 V 50 Hz circuits which are supplied by the working own-requirement and
reserve transformers.
In normal operating regime the 6 kV supply for own-requirement users is
provided by two 63 MV working own-requirement transformers at a voltage of
20/6.3-6.3 kV with two separated windings. The own-requirement transformer
has a dead connection between two switches in the unit generator circuit which
are switched in series. Each own-requirement unit transformer is connected to
two 6 kV working sections for supplying own-requirement users.
The fact that the circuit of each generator has two switches means that
the working own-requirement transformers can be used for starting up the unit
and for shutting it down if the generator circuit is defective, and that the
own-requirement power supply will be maintained if the unit is switched off
for technical reasons, and also in the event of any electrical breakdowns at
the reactor higher up than the generator switches, in particular, short-
circuiting in the unit transformers. %
The two-switch circuit is such that the run-down of the turbogenerator
can be used for supplying power to the feed pumps, which provide the water
supply to the core during the first 45 seconds from the beginning of a
design-basis accident in the event of loss of own-requirement power supply
from the higher-voltage (external-requirement) circuit. The 20 kV switches
are disconnected from the unit transformer by the separate operation of the
turbogenerators when they are running down.
In this case the generator voltage varies in proportion to its rpm by
means of a special "run-down unit" which is connected to the turbogenerator
excitation regulator; this "run-down unit" ensures that the rotor current in
the generator is maintained constant with the decrease in frequency.
The "run-down unit" is switched on when the design-basis accident
signal is given and the turbine shut-off valves close.
Redundancy is achieved with the own-requirement unit transformers by
means of a 63 MVA reserve transformer connected by an open-air line to the
330 kV outdoor switching station.
The 6 kV users of each turbogenerator are connected to the
corresponding sections of the own-requirement unit transformer, and the
reactor-part and whole-unit users are distributed evenly between the sections
of the two own-requirement unit transformers; the electric motors of
interchangeable (mutually redundant) mechanisms are connected to different
sections.
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2.10.3.3. Emergency power supply system
Users of the first and second groups receive power from the emergency
power supply system, the power sources for which, in addition to the working
and reserve own-requirement transformers, are independent (storage batteries
with static transformers and diesel generators).
The users of the first and second groups are subdivided into users from
safety-related process systems and "whole-unit users" for which a power supply
is absolutely essential, even when the plant's own-requirement power supply
has been totally cut off.
2.10.3.4. Circuit diagram for the 0.4 kV emergency power supply system
for the first group and for the direct current circuit for
safety systems
Safety system users of the first group include the isolating mechanism
for the accident localization (containment) system and hydrogen removal
system, the fast-acting valves and gate valves on emergency core-cooling
system (ECCS) lines and monitoring, protection and automatic control devices
of safety systems.
Three independent power supply sources (storage battery with static
inverter transformers and 6 kV and 0.4 kV own-requirement sections) are
foreseen for supplying power for users of the first group of each safety
sub-system.
The direct-current distribution panel of the safety sub-system receives
power from a rectifier connected to the 0.4 kV section of the emergency power
supply for the second group (NNBS),and when power is lost in this section from
the storage battery operating in the "buffer" regime.
Users of 0.4 kV alternating current of the first group are connected to
a 0.4 kV section (NNAS) which receives power from the direct-current
distribution panel through static inverter transformers.
In normal reactor operating regime the direct-current distribution
panel and the NNAS 0.4 kV section of each safety sub-system are connected to
the monitoring and control devices and automatic control systems of the
corresponding safety sub-system, and in design-basis accident regime they have
to cope with an additional load, that of the electrical drives of gate valves
and other valves of the ECCS and the accident localization (containment)
system. In order to prevent overloading of the inverter transformers above
the permissible levels with the current for starting up the electrical drives
for gate valves, the drives are actuated in stages following the design-basis
accident signal.
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2.10.3.5. Circuit diagram for the 0.4 kV whole-unit emergency power supply
system for the first group and for the direct-current ,circuit
Whole-unit users of the 0.4 kV emergency power supply system for the
first group are the "SKALA" central monitoring system, the control and
protection system, the dosimetric monitoring system, monitoring and measuring
instruments and automatic control systems of the reactor, turbine and
generator and the fast-acting pressure-reducing mechanism.
In order to supply power to users of the whole-unit emergency power
supply and direct-current system, there are two whole-unit emergency power
facilities, each of which include the following: power sources (storage
battery and static inverter transformers), direct-current distribution panel,
first-group 0.4 kV emergency power supply distribution panels and
own-requirement 6 kV and 0.4 kV sections.
The direct-current distribution panel of each whole-unit emergency
power supply facility receives power from a rectifier connected through a
6/0.4 kV transformer to the 6 kV section of the second-group emergency power
supply and, if power in this section is lost, from a storage battery operating
in the "buffer" regime.
First-group 0.4 kV users are linked to NNA sections through TKEO
thyristor commutator systems. NNA sections receive power through static
inverter transformers from the direct-current distribution panel.
Each user of the whole-unit circuit of the emergency power supply
system has two power sources. For the second source either the circuit is
used or another inverter transformer.
For users which cannot tolerate an interruption in power supply of more
than 10-20 ms (the "Skala" central monitoring system and control and
protection systems), the change-over to reserve power source is performed by a
TKP thyristor switching commutator, which changes the power supply to the user
over from one source to another in 10 ms. For users which can tolerate an
interruption in power supply of up to 100-200 ms, there are relay-contact
switching devices.
Systems for which redundancy is foreseen (the "A" and "B" feeder, the
"Skala" central monitoring system, the 1000 Hz and 400 Hz "Skala"
transformers, the emergency protection system control panel units, etc.) are
supplied by one of the emergency power supply facilities.
The power supply to systems for which redundancy is not foreseen
(control and measurement instruments, automatic control systems, regulation
systems, etc.) is provided by two emergency power supply facilities.
195
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
FOR OFFICIAL USE ONLY
Direct-current users (emergency protection system control panels,
warning systems, etc.) receive their power from both direct current
distribution panels. Switching over from one distribution panel to another is
performed manually.
2.10.3.6. Circuit diagram for the 6 kV and 0.4 kV own-requirement emergency
power supply systems for the second group
Users from the second group of safety systems are mechanisms of the
ECCS and the accident localization (containment) system.
Whole-unit users of the second group are mechanisms of the auxiliary
turbogenerator systems, certain auxiliary reactor systems (intermediate
circuit, cooling systems of the fuel cooling pond, blowdown and cooling
system, and so on).
For supplying power to users of the second group, there are three
emergency power supply sections of 6 kV and 0.4 kV (in accordance with the
number of safety sub-systems). Whole-unit users are distributed over the
safety sub-system sections.
Diesel generators with a capacity of 5500 kW were used as an
independent power supply for the 6 kV emergency power supply sections at the
fourth unit of the Chernobyl' nuclear power plant. The startup time of the
diesel generator is 15 seconds.
?The diesel generators take up the load in stages. The time for each
stage to be taken up is 5 seconds.
The diesel generator is started up automatically, with the load taken
up in stages, upon receipt of the design-basis accident signal or current loss
signal.
When one of these signals is received by the circuit for automatic
startup of the diesel generator with take-up of the load in stages, the
following commands are issued:
- Startup of diesel generator;
Switching off of both section switches linking the working
own-requirement 6 kV section with the emergency power supply
_section;
- Switching-off of the load on the 6 kV emergency power supply
section ("clearing of the section");
Blocking, by automatic stand-by startup of mechanisms connected to
the particular emergency power supply section.
196
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
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After the diesel generator has been started up and connected to the
section, automatic switching-on of the own-requirement mechanism switches
takes place in stages at 5-second intervals in accordance with the schedule of
load take-up in stages (Fig. 2.53).
Depending on the signal received, the circuit automatically switches on
in stages the corresponding mechanisms needed for a design-basis accident or
for loss of current by own-requirement users.
For the 0.4 kV users of the second group there are 0.4 kV sections of
the safety systems (NNBS) and whole-unit 0.4 kV sections which are independent
of them. Each section receives power from the corresponding 6 kV section of
the safety system through a 6/0.4 kV transformer.
s
The number of NNBS 0.4 kV safety system sections corresponds to the
number of operating safety sub-systems.
Emergency power supply 6/0.4 kV transformers for the second group are a
stage of load take-up by diesel generators that cannot be switched off.
No mutual redundancy is foreseen between the 6 kV and 0.4 kV sections
of the second group since there is redundancy of the users themselves.
2.10.4. Controlling safety systems
The controlling safety systems are designed to switch on automatically
devices of the protection, localization (containment) and power-supply safety
systems and to monitor their operation.
For each of the three safety sub-systems there is an independent
controlling safety system.
A controlling safety system issues the design-basis accident signal if
the pressure in the containment, lower water line or drum separator enclosures
rises to 5 kPa with confirmation of the decrease in the level of the separator
by 700 mm from the nominal level or decrease in the gradient between the
pressure header of the main circulation pumps and the drum separator to
0.5 HPa.
In order to increase their reliability, all three controlling safety
systems have been constructed independently from one another, i.e. each of
these controlling safety systems has its own engineered structures and
electricity supply, and separate areas for engineered structures and cable
conduits. Four sensors are provided for ,issuing the signal indicating high
pressure in the containment, drum separator and lower water pipe enclosures.
if two or more of the sensors are triggered, a signal is issued.
197
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FOR OFFICIAL USE ONLY
The signal indicating a pressure decrease in the drum separator and
also the signal indicating a decrease in the pressure gradient between the
pressure header and the drum separator are issued when any two of the sensors
provided are triggered.
The design-basis accident signal is issued independently for either
half of the reactor.
When the design-basis accident signal is issued, the controlling safety
system issues instructions for output Actions for switching over the
corresponding safety system devices and for switching on the diesel generators
and mechanisms for taking up the load in stages.
The design provides for the possibility of remote control of the safety
system, for which the operational part of the reactor control panel has
control switches for each controlling safety system.
In this case the half of the plant in which there is an accident is
selected automatically, for which use is made of the information part - which
is independent of the controlling safety system - from the emergency
protection system triggering circuit for process reasons.
For purposes of monitoring the correct functioning of the controlling
safety system there are warning lights and acoustic signals indicating that
Instruments are defective.
The safety system is monitored and controlled from the safety panels
set up in the area of the operational circuit of the reactor control panel and
on a redundant control panel.
The safety panels contain devices for controlling the pumps of the
emergency core cooling system (ECCS), the accident localization (containment)
system, the safety system equipment, instruments for monitoring the flow of
ECCS water into the reactor, etc.
Figure 2.54 shows a flow chart of the controlling safety system.
198
FOR OFFICIAL USE ONLY
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0
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73
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VINO ASfl IVIDII.10 /103
Cooling 1 1
service
water
004 04
Fig. 74i,
12
Schematic drawing of the reactor's emergency cooling system.
1. Reactor 2. Steam separators 3. Suction header 4. Main circulation pump
5. Pressure header 6. Pressure suppression pool 7. ECCS vessels
8. ECCS pumps for cooling the damaged half of reactor 9. Heat exchangers
10. Clean condensate container 11. ECCS pumps for cooling the undamaged
half of the reactor 12. De-aerator 13. Feed pump.
/LINO ISII IVIDIA10 1103
0
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EINO aSfl rIVI3I330 1103
n.)
ECCS
t/sec
1.0
4 5
4N.
Required supply of water to the damaged half of the reactors in different accident situations
or.
1
2
0 20 40
160
120
100
80
60
20.
GGrp.hdr.
, t/h
ECCS
0
40
60
SO
/Change of scale
1X
Time (secs)
60 ao
100
120
Type of accident: 1. Rupture of a 04 900 me pressure header
2. Rupture of a Du 300 me down pipe
3. Rupture of a group header as far as the non-return valve
. Fig! 2.44
160
>80
(800
5600
Time (secs)
AINO USTI IVIDI130 110.1
Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
AINO Isa IVIDIII0 110,1
Fresh steam from separators
SYSTEM FOR DISCHARGING STEAM FROM THE MAIN SAFETY VALVES INTO THE PRESSURE SUPPRESSION POOL
???."
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?
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Header_804, J77 411
V qitpkOre.aker_.
e., ? a?
II I III II
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*/ NSV = main safety valve.
-
4,5
Fig. 2.45.
0
?117
turbogenerators
leaktight compartment
Level 'mark
.eRESAURE_SUPPRESSION POOL
A Y Y
4: 4 Level mark
KINO HS11 rIVIDLE,I0 110,1
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
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41'
M
(Ii
7? z_
ii.dit i?
I
.1 III
Es
202
FOR OFFICIAL USE ONLY
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
FOR OFFICIAL USE ONLY
203
FOR OFFICIAL USE ONLY
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
VINO 1Sfl IVIDIA30 Hal
NPA
4 2
44
0
Fig 2.48 Pressure changes in the buildings of the second unit of the Smolensk nuclear power station
and the fourth units of the Kursk and Chenobyld nuclear power stations during pressure header
failure
Design diagram of buildings
tsi
iamarimum
nummuirr
mrismusaw
prirPtag
c
(3
iO
2.4P
SO
100
/SO
Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
if:Mflci Npa
4 .1 di f,6
Design diagram of buildings
1
631
0
Fig
2.49
I0 150.
'00 (10
Pressure changes in the buildings of the second unit of the Smolensk ,
nuclear power station and the fourth units of the Kursk and Chernobyl
nuclear pwer stations during failure of the disturbing group header
eCt2
eS0 C)
VINO Isa IVIDLIA0 110.1
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
VINO asa 110.4
Vga
V?
. I
-
3
10 0 ? Total entry of hydrogen
Entry of H2 into the accident confinemeqt area
Volume of 112 entering the area
Entry of H2 accumulated in coolant before accident
*/
Entry of 112 owing to water?zirconium reactor?
Entry of H2 owing to water radiolysis
90
80 -70 -
CO -
50 -
40
30
10 ...
According to existing
standards the percentage
of. reacting zirconium is
not more than 1% of the
clodding mass of fuels from
the channels in one distrib?
uting group header. This
graph gives the entry of ?
hydrogen on the basis of 3%
reacting zirconium (design
safety)
s
.10'
Fig 2
24 4d
Z,Iol
h
ur
AIN() ISfl IVIDIA10 110,1
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LEGEND TO FIG 2.51
HYDROGEN REMOVAL SYSTEM
1. Nitrogen supply
2. Discharge to vent flue via V--1, A
3. Discharge into GARS or to vent flue without GARS
4. To trap
5. Mater outlet - water inlet
6. Air suction
7. Air suction
8. Sub-Syste? No. 1
9. Sub-System No. 2
10. Sub-System No. 3
11. Stea?-gas discharge co?ponent
12. Downcomer shaft
13. Pressure suppression pool
14. Downcomer shaft
15. Discharge for gas analysers
16. Neasure?ent of hydrogen concentration
207
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4.1
0
I.
40
0
Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
FOR OFFICIAL USE ONLY
IM?
208
FOR OFFICIAL USE ONLY
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FOR OFFICIAL USE ONLY
Key to Fig. 2.52 [This figure has been translated to the extent that the
quality of the original permits]
Figure caption: Own-requirement circuit diagram for the fourth unit at the
Chernobyl' NPP
1. 63 MVA 350/6.3-6.3 kV reserve transformer
2. To 330 kV outside switching station
3. 63 MVA 20/6.3-6.3 kV working own-requirement transformer
4. Unit transformer
4a. To 750 kV outside switching station
5. 20 kV
6. 6 kV reserve busbar
7. 6 kV sections of the third emergency power supply group
8. 7RB
9. 7RA
10. 8RB
11. 8RA
12. Main circulation pumps
13. First-stage condensate pump
14. Second-stage condensate pump
15. Feed pump
16. Other mechanisms
17. Working own-requirement transformers
18. Other own-requirement transformers
19. To 6 kV section of the third unit
20. 6/0.4 kV reserve transformer
20a. 0.4 kV reserve busbar
21. 6 kV sections of the third emergency power supply group
22. Diesel generator 2
22a Diesel generator 1
23. Diesel generator 3
24. 7RNB
25. 7RNA
26. 8RNA
27. Process water pump
28. Non-accident-half ECCS pump
29. Accident-half ECCS pump
30. Emergency protection system channel cooling circuit pump
31. Sprinkler-cooling system pump
32. Emergency supply transformer 82 TNP
33. 0.4 kV sections of the second emergency supply system group
34. 2NNBC
35. 72NNB
36. 71NNB
37. 92NNB
37a. 91(?)NNB
38. 3NNBC
209
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39. 0.4 kV users of the second safety system
40. 0.4 kV users of the first safety system
41. Electric motors of ventilation systems
42. Whole-unit users
43. Whole-unit emergency supply system facilities:
44. Emergency supply system facility for second safety system (SD II)
45. Emergency supply system for first safety system (SB I)
46. First
47. Second
48. Emergency supply system facility for third safety system (SB III)
49. Monitoring and control instruments, automatic control systems,
regulators
50. Emergency protection system and "Skala" central monitoring system users
51. Electric drives of gate valves
52. 21NNA
53. 22NNA
54. NNB
55. 11NNA
56. 12NNA
57. NNB
58. 11NNA
59. 12NNA
60. 13NNA
61. 14NNA
62. 21NNA
63. 22NNA
64. 23NNA
65. 24NNA
66. 0.4 kV sections of first emergency power supply group
67. 2NNAS
68. 1NNAS
69. 3NNAS-1
70. 3NNAS-2
71. Electric motors of isolating mechanism and ECCS
72. Automatic control, protection and regulation devices
73. 2 MPS
PTS-63
74. 2 VUS
75. 1 MPS
PTS-63
76. 1 VUS
77. 1 VU
78. 11 MP
79. 12 MP
80. 13 MP
81. 14 MP
82. 1 VUP
83. 91 NNB
210
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84. 21 MP
85. 22 MP
86. 23)11'
87. 24 MP
88. 2 VU
89. 2 VUP
90. 91 NNV
91. 3 MPS-1
PTS-63
92. 3 MPS-2
PTS-63
93. 3 VUS
94. Direct current distribution panel 3 S (3ShchPTS)
95. Direct current distribution panels
96. =232V
97. Direct current distribution panel 2 S (2 ShchPTS)
98. =232V
99. Direct current distribution panel 1 S (1ShchPTS)
100. Direct current distribution panel 1 (1ShchPT) = 253 V
101. = 232 V
102. Direct current distribution panel 2 (2ShchPT) = 253 V
103. = 232 V
104. 232 V
105. Direct current users
106. 3ABC
SK-16
108 elec.
107. lABC
SK-16
108 elec.
108. lAB
SK-52
118 elec.
109. From 108 elec.
110. Emergency power supply transformer 2 S (2TNPS)
111. Emergency power supply transformer 1 S (1TNPS)
112. Emergency power supply transformer 3 S (3TNPS)
113. Emergency power supply transformer 72
114. Emergency power supply transformer 73
115. Emergency power supply transformer 71
116. Emergency power supply transformer 93
117. Emergency power supply transformer 92
118. Emergency power supply transformer 91
119. Emergency power supply transformer 81 [?]
120. Transformer 225 T
211
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Declassified and Approved For Release 2013/07/15
0-1-0009Z001-0001L6600-60dCll-V10
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
FOR OFFICIAL USE ONLY
Key to Fig. 2.53
Horizontal captions (see column numbers on figure)
1. No
2. Name of emergency power supply user
3.-13. Starting data
3. Number attached to one safety sub-system
5.-6. Power (kW)
5. Rated
6. Consumption
7. Nominal current, In (A)
8. Startup current, It (A)
13. rpm
1A.-17. Theoretical values
1/!. Theoretical power (kW)
15. Startup power (kW)
16. Slippage (Snom)
17. CosS startup
18.-32. Design-basis accident regime
19.-29. Switching stages
18. Number of motors started up
19. 15 s I
20. 20 s II
21. 25 s III
22. 30s IV
23. 35s V
24. 40s VI
25. 45 s VII
26. 50 s VIII
27. 55 s IX
28. 60s X
29. 65 s XI
30.-32. grated (kW)
'rated (A)
30. Per stage
31. Up to 10 min
32. Up to 30 min and beyond
33.-47. Loss of current by own-requirement users
33. Number of motors started up
34.-35. Switching stages
34. 15 s I
35. 20s II
36. 25 s III
37. 305 IV
38. 35 s V
213
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39.
40.
40s VI
45 s VII
41.
50 s VIII
42.
55 s IX
43.
60s X
44.
65s XI
45.-47.
Prated (kW)
'rated (A)
45.
Per stage
46.
Up to 10 min
47.
Up to 30 min
48.
Notes: [opposite
vertical columns
12,
13:]
Not started up in stages
Caption at bottom ri&ht of figure
This table was compiled for one safety sub-system. The tables for
the other two sub-systems are similar.
Vertical captions (see line numbers on figure)
1. Transformers for emergency supply system and plant's own requirements
2. Process water pump
3. Non-accident-half ECCS pump
4. Sprinkler-cooling system pump
5. Accident-half ECCS pump
6. Emergency protection system channel cooling circuit pump
7. Reserve
8. Clean condensate pump
9. Emergency feed pump
10. Fire pump
11. Emergency supply system transformer
12. Circuit cooling pump
13. Tank pump
214
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414.
s
foOparcur l?
;my
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0 IN
VC:
t?kil
Pah.
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rk
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2.11. Other safety-related systems
2.11.1. Multiple forced circulation circuit (MFCC).
A description of the multiple forced circulation circuit (primary
coolant circuit) and its main components is given in sections 2.6 and 2.7.
2.11.2. CPS channel cooling system
The system for cooling the CPS (control and protection system)
channels is designed to ensure the requisite temperature condition-a- for these
channels together with the control elements and servodrives of the CPS.
The system performs the following functions:
- It maintains a temperature of 400 at the coolant water inlet
to the control channels;
- It removes a thermal output of 2&.1 MW from the channels for the
CPS control elements and servodrives;
- It ensures cooling of the CPS control element and servodrive
channels at a nominal flow rate for -6 minutes when the pumps
are not functioning;
It maintains a nonexplosive concentration of hydrogen under all
working conditions;
- It maintains the requisite amount of water cooling the
and CPS servodrives;
channels
- It ensures that there is emergency protection of the reactor if
the cooling system is disrupted.
These functions are performed with allowance for single failure in
the system of an active element or a passive element with moving mechanical
parts.
A schematic diagram of the CPS channel cooling system is shown in
Fig. 2.55.
The system constitutes a circulation loop operating by gravity.
Water from the top emergency supply tank flows by gravity into the
pressure (distributing) header and is distributed through the channels. The
channels contain the elements of the control and protection system and tubes
containing the fission chambers and power density monitors. Some of the
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channels are used to control the flow of water cooling the graphite of the
lateral reflector.
After flowing through the CPS channels the cooling loop water
transfers,its heat to the service water in the heat exchangers of the system.
Dependineon the temperature of the service water and degree of contamination
of the heat exchange surface, the required heat removal is ensured by two
heat-exchangers, two others being redundant.
After the heat-exchangers the water flows into the lower tanks of the
system, in which there is automatically maintained a level ensuring stable
functioning by the pumps under steady-state and transient conditions. The
total volume of the lower tanks is such that they can hold all the water in
the system to be received if the pumps stop.
There are four pumps for feeding the water from the lower tanks to
the emergency supply tank. The delivery rate for each pumps is ?700 t/h at
a pressure head of 0.9 MPa. Two of these are in operation, 'while two are
redundant.
Provision is made for reducing the probability of all pumps failing
for the same reason (the pumps are located in different rooms, have
independent power supplies and so forth).
The output of the working pumps exceeds the throughput of the cooling
system, hence some of the water is always being discharged from the emergency
supply tank into the lower tanks of the system (the level of the water in the
emergency tank is kept at the overflow mark).
Radiolysis in the reactor core causes the generation of hydrogen from
the water in the CPS cooling system.
To prevent the formation of an explosive concentration of hydrogen
there is continuous ventilation of the space above the water in the top and
bottom tanks, together with monitoring of the hydrogen content in the CPS
cooling water as well as in the space above the water in its tanks.
The emergency water supply tank is connected to the atmosphere by
four breather pipes which lead off from the top of it. In addition to this
the space above the water in the tank is constantly blown through with
compressed air. If there is a failure in the compressed air supply system,
the space above the water is ventilated by air ejection, using the excess
water continuously drained from the emergency tank into the lower tanks
through the overflow. If the system stops, all the water from the emergency
tank is drained into the lower tanks.
The lower tanks of the system are continuously blowndown with
compressed air and they also receive the air ejected from the emergency tank;
furthermore, since the tanks are at reduced pressure, they receive air from
the building through a special iirie With a Valve. Reduced pressure in the
tanks and removal of air blown through them is ensured by a special tank
ventilation system.
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If any of the blowdown (ventilation) systems should fail, those still
functioning keep the hydrogen concentration at a safe level.
To maintain the requisite quality of the water in the cooling system,
it has to be constantly purified. Water is fed for clean-up from the pump
pressure header and returned to the lower tanks.
If there are disruptions in the cooling system (reduced level in the
emergency supply tank or a fall in the water flow rate), signals are sent for
emergency shutdown.
Parameter monitoring and control of the CPS channel cooling system is
carried out by the operating staff from the unit control panel. The system
was thoroughly checked during the startup and adjustment operations and during
operation of the unit.
2.11.3. Blowdown and cooling system
The blowdown and cooling system shown in Fig. 2.56 is intended to
cool the blowdown water of the MFCC bled off for clean-up, followed by
reheating before it is returned to the MFCC under nominal conditions, and to
reduce the temperature of the circulation water to the required level under
cooling conditions.
Under nominal conditions the MFCC coolant flowing at 200 t/h (100 t/h
from each loop) is pumped by the main circulation pumps to a regenerator where
it cools from 2850C to 680C through heat removal to a cold counterflow,
and is then further cooled down to 500C by the water of the intermediate
circuit in the blowdown afterheater, from where it enters the circuit water
clean-up system. As it passes through the regenerator in the opposite
direction, the cleaned water heats up from 500C to 2690C and is recycled
to the steam separators through mixers in the feed water piping. It should be
pointed out that either of the two after-heaters in the blowdown and cooling
system may operate in this mode.
When cooling the unit the blowdown and cooling system reduces the
temperature of the water in the MFCC, starting at 1800C, down to the
temperature required for repairs to the unit. Circulation takes place along
the line: steam separators - cooling pumps - larger after-heater - steam
separators.
The blowdown and cooling system may also be used to remove residual
heat from the reactor when there loss of current for the power unit's own
needs. In this mode the operational system is the same as for the cooling
mode.
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2.11.4. Gas circuit system
The main layout of the system is shown in Fig. 2.57.
Under nominal operating conditions the gas circuit system works in the
following way: the nitrogen-helium mixture, when it emerges from the reactor,
passes through the fuel channel integrity monitoring system, where a
channel-by=channel temperature check is made and the moisture content of the
nitrogen-helium mixture is monitored for groups of channels.
Having cleared the fuel channel integrity monitoring system, the
mixture passes through a series of condensers, air heaters and filters in
which iodine vapours are deposited, and reaches the compressor intake of the
helium scrubbing unit, in which apparatus hydrogen, oxygen, methane, carbon
dioxide, carbon monoxide and ammonia impurities are removed from the mixture,
down to a concentration permitting normal reactor use.
Removal of radioactive argon-41 takes place in cooling tanks fat
-1950C].
After passing through the scrubbing system, the mixture is returned to
the reactorpile. A hydraulic seal is fitted to the pipe which introduces the
mixture into the pile to prevent the pressure from rising above permitted
'levels, i.e. higher than 1-3 kPa.
In order to reduce leakage of helium from the reactor pile, nitrogen
(99.99997. pure) is introduced into the metal structures of the reactor at a
pressure of 2-5 kPa. A hydraulic seal is fitted to the feed pipe.
In the gas circuit system there exists the possibility of flushing the
reactor pile with nitrogen. In this event the nitrogen is dumped via the
activity reduction system.
In the gas circuit system measurements are made of flow rate, impurity
concentration, moisture content, temperature and pressure of the
nitrogen-helium mixture, and the circuit is monitored for radiation. All
results are displayed on the gas circuit control panel.
The system is controlled from the gas circuit control panel.
2.11.5.Cooling of spent fuel storage ponds
The pond cooling system is designed to stabilize the temperature of the
water ponds, which is heated by the decay heat from the spent fuel, in all
operating modes including a total power failure affecting in-house
requirements. The system maintains the temperature of the water in the
cooling ponds:
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- Under normal operating conditions - at no more than- 50?C, where
the maximum decay heat is 1800 kW;
- During simultaneous unloading into the pond of 5% of the fuel
assemblies from damaged channels - at no more than 700C, with
maximum decay heat in the pond of no more than 3000 kW;
- When heat removal ceases as a result of any departures from normal
operating conditions or loss of electric power to the system
(- the water temperature should rise to no more than 80?C in the
20 hours after heat removal has ceased.)
During this time measures must be taken to restore the functioning
capacity of the system.
Water quality in the cooling ponds is maintained by a bypass
purification system. Arrangements exist to exclude the possibility of
accidental drainage of the ponds. The space above the water is ventilated.
The main layout of the cooling pond system is shown in Figure 2.58.
The cooling of the ponds is achieved by means of a closed loop. Water,
heated in the ponds by decay heat production, passes from the upper regions of
the ponds to heat exchangers, where the heat is transferred to service water.
The required heat reduction is achieved with one heat exchanger, while a
second stands in reserve. Having passed through the heat exchangers the water
is returned to the pond by one of two pumps at a rate of 160 m3/hr and
at a pressure of - 20 m water column (the second pump is a reserve pump).
The cooling water outflow and return pipes are so arranged that if they
were to rupture the level in the ponds would not fall below the permissible
minimum.
To prevent overfilling of the ponds each has an overflow.
To prevent the formation of an explosive concentration of hydrogen in
the space above the water in the ponds, constant ventilation is provided by
air taken from the central hall. Should the pond ventilation system suffer a
malfunction, the flow cross-section of the ducts which connect the ponds with
the central hall is such that one may view them as a single compartment with a
volume of 40 000 m3, ventilated by an independent ventilation system.
The water in the ponds is purified by means of a loop which is
independent of the cooling system.
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The following parameters are monitored:
- Water levels in the ponds;
- Temperature of water in the ponds;
- Flow rate of cooling water etc.
Operating personnel control the system and monitor its process
parameters from the unit control room.
The system underwent comprehensive and direct testing against design
specifications during commissioning operations and while the power unit was in
use.
2.11.6. Ejection cooling system
The ejection cooling system (Fig. 2.59) is designed to remove heat from
the leaktight compartments. For each of the two leaktight compartments there
are four groups of coolers set at level 5.0 in the housing of the main
circulation pump tanks. Each group consists of four coolers, and each cooler
has a capacity of 2500 m3/hr. Each group has an independent air supply. As
regards their water supply, the coolers are divided into two independent
sub?systems of eight coolers each and connected to different feedback Sylphon
pumping systems.
Air at maximum temperature is drawm off from the upper regions of the
downcomer shafts through four pipes, fed to each group of coolers where it is
cooled by jets of water down to 350C in summer, and to 180C in winter, and
then passes into the compartment containing the main circulation pump tanks.
The cooled air removes heat from the mechanical equipment and unplanned
coolant leakages. To prevent escape of dispersed moisture with the air,
separators are installed at the outlet from the coolers. Apart from cooling
the air and eliminating excess moisture, the ejection coolers also remove
aerosols, including radioactive iodine.
The ejection cooling system is compact and contains no active elements
which require maintenance or control while it is operating.
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2.11.7. Radiation monitoring system
The nuclear power station radiation control system is a component part
(sub-system) of the automated station control system and is designed to
gather, process and present information on radiation conditions in the station
compartments and outside, on conditions in the process media and circuits and
on irradiation doses to personnel and individuals from the population in
accordance with the norms and statutes in force.
The radaition monitoring system as a whole can be divided into two:
the process radiation monitoring and the radiation dosimetry systems. The
purpose of the process radiation monitoring system is process optimization,
and also to monitor the condition of the protective barriers against the
spread of radionuclides. The purpose of the radiation dosimetry system is to
monitor the radioecological factors arising from the operation of the facility
and, in the final account, to determine the internal and external irradiation
doses received by staff and individuals in the population.
The off-site dosimetric monitoring system is distinct from the
radiation dosimetry system, and:
- Determines the activities and nuclide compositions of radioactive
substances in the atmosphere;
- Monitors gamma radiation dose exposures in the area;
- Monitors radioactive fallout;
- Monitors ground water activity in test bore-holes;
- Determines the content of radioactive substances in soil,
vegetation, locally-produced feedstuffs, food products and so on.
A structural diagram of the radiation dosimetry system is shown in
Fig. 2.60.
The following are used for radiation monitoring:
(1) The combined AKRB-06 unit, which includes detection units and
equipment, informtion processing equipment, surface contamination
monitoring units and units and dosemeters for monitoring the
station personnel;
(2) Individual, portable and wearable devices;
(3) Laboratory equipment and instruments.
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The radiation monitoring structure takes the form of a data and
measurment system with a large number of dispersed information sources and
capture devices, arranged in a radial-annular manner; under this system, the
detection units, the UNO-06r information storage and processing unit and the
local units for signalling when established values are exceeded are linked
radially. The UNO-06r system's internal links and links with the UNO-01r
monitoring and data exchange device are arranged in a ring.
The AKRB-06 thus monitors continuously the readings from the detecting
units and devices, transmits information on all channels to the computer,
signals failures of its components and controls the shut-off equipment on the
sampling lines. The devices for displaying information (display, console,
signalling units) are located on the radiation monitoring board.
The detection units feeding into the AKRB-06 measure:
The gamma exposure rate within the range 10-5 to 103 R/h
(BDMG-41, BDMG-41-01, UDMG-42, UDMG-41-02);
The activity concentration of gamma emitters in liquid process
media and circuits within the range 5 x 10-11 to 10-3 Ci/L
(UDZhG-04r, UDZhG-05r, UDZhG-14r1);
The activity concentration of iodine vapours in air within the
range 10-11 to 10-6 Ci/L (BDAD-06);
- The activity concentrations of aerosols with dispersion phases
containing beta emitters within the range 10-13 to 10-9 Ci/L
(BDAB-05);
The beta activity concentration of inert gases in the air and
process media within the range 10-9 to 1.4 x 10-4 Ci/L
(UGDB-08) and 10-5 to 0.3 Ci/L (UDGB-05-01);
- The activity of long-lived beta-emitting aerosols in the
gas-aerosol releases to the vent stack within the range
3 x 10-14 to 3 x 10-10 Ci/L;
The activity of short-lived beta-emitting aerosols in the
gas-aerosol releases to the vent stack within the range
1.5 x 10-12 to 1.5 x 10-8 Ci/L;
The activity of beta-emitting inert gases in gas-aerosol releases
to the vent stack within the range 8 x 10-9 to 8 x 10-5 Ci/L;
- The activity of the gamma-emitting vapours in the gaseous phase of
the gas-aerosol releases to the vent stack within the range
3 x-1013 to 3 x 10-10 Ci/L.
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Gas-aerosol releases to the ?vent stack are measured using RKS-03-01
and RKS-2-02 radiometers. The airflow through the vent stack is measured
using a partial flowmeter with a metal-polymer sensing element. The
measurement data are processed by a computer.
Each power station unit has a total detector complement of 490 units,
of which approximately. 400 are in production areas with a continuous or
restricted staff presence.
The surface contamination monitoring unit notifies staff when
contamination exceeds the following established threshold levels:
- For the skin of the hands: beta emitters within the range 10 to
2000 counts/min./cm2 (RZG-05-01, SZB-03, SZB-04);
- For the skin of the body or basic protective clothing, beta
emitters within the range 5 to 2000 counts/min./cm2 (RZB-04-04);
- For means of transport, on leaving the station in order to detect
objects to be investigated in detail using other means, gamma
radiation within the range 2.78 x 10-2 to 0.278 pR/s (RZG-05);
- For personnel, on leaving the station, for detection and
subsequent detailed investigation, gamma radiation within the
range 1.4 x 10-2 to 0.14 pR/s (RZG-04-01).
The personnel irradiation monitoring unit continuously monitors
external irradiation. To this end, the basic items of equipment used are:
- Sets of individual dosimetric photomonitors to measure the total
exposure to gamma radiation within the range 0.05 to 2 R at
energies of 0.1 to 1.25 MeV (IFKU-1);
- Sets of thermoluminescence dosemeters to measure exposures to
X-ray and gamma radiation in the energy range 0.06 to 1.25 MeV
within limits of 1.0 to 1000 R and 0.1 to 1000 R (KDT-02);
- Gamma exposure rate dosemeter-indicators in the range 0.1 to
9.9 R/h; there is also a range of other dosemeter variants with
similar characteristics.
The internal irradiation recording equipment measures the whole-body
burden of 137Ce and "Co nuclides and the thyroid 1311 burden (MSG-01).
In addition, DGDK-type semiconductor detectors and their analysis and
processing equipment are used to identify a range of radionuclides in the
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human body. There is a wide range of portable and wearable dosemeters and
radiometers in the total stock. For example, the following dosemeters are
used:
- For measuring exposures to gamma and X-ray radiation in the energy
range between 15 keV and 25 MeV with a measurement range between
0.1 pR/s and 11 R/s (DRG, DKS);
- For measuring neutron dose equivalent rates between 0.05 and
5000 prem/s (KDK-2);
- For express measurements of the specific activities of samples
(RKB4-1 beta radiometer) within the range 2 x 10-12 to
10-7 Ci/L;
- For measuring nuclide activity concentration in liquids and air
for alpha, beta and gamma radiation over various energy ranges
(RZhS-05, RGA-01, MKS-01 and others).
Off-s!te dosimetric monitoring is carried out in the area of the
station within a radius of approximately 35 km. It is carried out by the
off-site dosimetry service, of the plant and is designed to obtain the
information required to evaluate the external and internal doses to
individuals in the population. The monitoring equipment is located at 38
posts, and inpludes total gamma dosemeters, vials for collecting atmospheric
fallout and seven aspiration sets.
Samples are analysed using semiconductor detectors, spectrometers and
analysers with microcomputers. On the basis of the data on releases into the
atmosphere through the power station vent stack and by means of automatic
measurement of the meterological parameters, a forecast is made using a
microcomputer of the radiation situtation in the power station area.
2.11.8. NPP control
The NPP is controlled on two levels:
As a station;
By unit. (See Fig. 2.6.1 "Basic structural diagram of station
control sytem"). Control over all plant safety systems is carried
out at unit level.
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Station-level control
At the station level, operational control is effected from the central
control board. At the station level, the operating staff are responsible for:
Control over the electrical equipment in the main electrical
connection circuit (750-kV line interrupters, unit transformers,
autotransformer and so on, 20 kV generator interrupters, 330 kV
autotransformer interrupters and interrupters for the 6 and 330 kV
back-up medium voltage transformers);
- Distribution of active and reactive power;
- Co-ordinating the work of the operating staff at the unit control
panels and in separate installations on the site.
On the central control board there are remote control keys for the
aforementioned interrupters, and also audible and visual signalling of
accident and fault conditions, and visual signalling of the condition of the
switching apparatus (interrupter in or out) on the mimic diagram.
The protection relay equipment, 'anti-accident automatics .and
telemechanics are housed in: the protection relay buildings :-for the
corresponding 750 kV and 330 kV distribution equipment; Microchip integrated
circuits form the basis of the 750 kV line protection relay equipment; they
monitor the function of each separate channel and make it possible to. test
them. Mass-produced electromechanical relays are also used in the protection
relay control devices and anti-accident automatics.
Unit-level cpntrol
The process installations and structures given unit are
controlled at unit level:
- the reactor and its supply facilities (main circulating pumps,
feed pumps, emergency feed pumps and so on);
- turbogenerators and auxiliary equipment;
- normal and back-up medium voltage supplies and so on;
- separate installations on the site: diesel power station, process
water supply pumps and so on.
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The above are controlled from the unit control board, which includes
the control board and display. The operator control circuit of the unit
control panel is split into control areas:
reactor control;
steam generator set control;
turbine, generator and medium voltage supply control.
On the operator circuit of the unit control board are located the
operators' work stations and the control panels for the:
senior reactor control engineer;
senior unit control engineer;
senior turbogenerator control engineer.
These control panels contain the following:
control apparatus;
monitoring system instruments;
?"Skala"
. units;
central monitoring system, call-up devices andldisplay
communications apparatus.
On the unit control panel operator circuit board are the following':
reactor channel mimic board;
CPS mimic board;
mimic diagram of the thermal and electrical parts of the unit;
- individual instruments for the monitoring system signalling
equipment.
The "Skala" central monitoring system monitors the main bulk of the
parameters. The most important parameters required for correct process
operation also have their own individual monitoring instruments. These
include instruments showing reactor output, drum separator level and pressure,
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
FOR OFFICIAL, USE ONLY
steam flow rate ex drum separator, feedwater flow rate into drum separator,
measurements from the physical power distribution monitoring system and the
CPS and so on.
Electricity is supplied to the unit control panel and the "Skala"
central monitoring system from the secure power supply so that even when there
is a power loss at the medium voltage buses, the operator does not lose
information on the conditions of the process parameters. On the unit control
panel operating circuit are 'located the controls for the process protection
and monitoring devices. Overall operative control of the unit is carried out
from the controller's board, which has telephone apparatus and loudspeaker
links.
Included in the operator circuit of the unit control panel there are
in addition special safety panels for each of the three safety sub- systems;
on these the back-up medium voltage power supply (diesel generators) and the
5
emergency reactor cooling and accident containment systems are controlled and
monitored.
A back-up control board is provided for the eventuality that the
reactor cannot be shut down and maintained in sub-critical condition from the
unit control board. On the operator circuit of the back-up control board are
located control panel, operator circuit panels and safety panels. On the
control panel are located the AZ-5 emergency protection system button, CPS
coupling disconnection switch, signalling board and so on. On the operator
circuit panels are the recorders for neutron power, drum separator pressure
and so on. The safety panels of the back-up control board are analogous to
those on the unit control board.
Local control panels are provided for the range of systems which
operate independently of the main processes; these include the gas circuit,
active waste treatment, radiation monitoring system, ejector gas sorption
scrubbing unit and the turbines.
Local boards are also provided for a range of units involved in the
main process (main circulation pump, electric feed pump, emergency electric
feed pump and so on), and these are installed with the equipment itself.
229
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Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
/LINO aSfl 1VIDI330 1103
N.
Compressed -
air feed
ti1 1 1 1 1 1 1 1 1
IT
1 I I I
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600
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II
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blowdown
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Cooling service water
6
1 1 1 500
<
Schematic representation of CPS fission chamber, power density monitor and failure detection
channel cooling system: (1) Reactor; (2) CPS, fission chamber and power monitor channels;
(3) Failure detection channels; (4) CPS heat-exchangers; (5) Drainage tank; (6) Circulation
tank; (7) CPS pumps; (8) CPS emergency tank; (9) Bypass cleaning unit; (10) Filters.
AINIO ASfl IVIDI330 1103
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To steam separators of the
second half of MFCC
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Fig.
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ii
IL
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From steam separator of
second half of MFCC
IF
xc
41 Feedwater from
41 electric feed pump
\
bs 1, To the
mehet_i> pressure-suppression
pool chamber
Active ventilation
Active drainage pipework
Fig. 2.57 Gas Circuit System.
Supply of nitrogen to the
metal structures of the reactor
Nitrogen supply
Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
VINO 3S11 lVIDkhIO 110.1
Filling and lo ventilation Filling and
make?up system make?up
rm????? ??alm........0??????????
Maximum-level
Overflow_fpinimum level
j Return of
ommood
i//purified water
4
, ? .
: ? ?
Fuel `'--?
storage
pond
?
,
????? ?
??'
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:
Removal of water
for purification
Im?????./.00. 0?1??=?????rd
Maximum level z
To ventilation
system
11
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ran ? ? ? J
Minimum level)" Overflow pt.
. ?
?
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storage J
. pond
Filling and
make?u
sylem
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Maximum
? Return of
purl le
water
Removal of water for purification
To ventilation
Removal of water
for purification
Fig. 2.58 System for Cooling Spent Fuel Storage Ponds
VINO 3S11 TVIDII30 110,1
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Water from the
t>
,134
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Layout of the Ejection Cooling System.
Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
VINIO HS11 IVIDIA30 1103
ts)
Radiation monitoring system
]Process radiation monitoring
I _
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Off?site dosimetric monitoring
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IRadiation monitoring
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A'INO S11 `IVID13,10 1103
Declassified and Approved For Release 2013/07/15: CIA-RDP09-00997R000100260001-0
UW,Y Dept?. titmothoetiox;
(1)
(1) Central Control Board
II construction phase
Chernobyl' NPP
(2)Unit controllboard Unit III
(2)
III 34 in"
Lam!~ immam**4 fturmaiwilem
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owl
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(4) Back-up contrq Board Unit III
Fig' Puc. 2.61
(3) Unit control board Unit
(3) 6W,4N 314
IV
speiva okra Ikgri
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OD
Aft
PW,4
tik01.
(5) Back?up control board Unit IV
humminutivoilit talip.ftpula
gum'. vryslaktwit
Basic structural diagram of station control system
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2.12. Reactor and unit operating modes
2.12.1.Normal operating modes
The reactor and unit operating modes can be divided into normal
operating modes and transient modes associated with equipment failure. Normal
operating modes consist of unit startup and shutdown, unit operation at power,
reactor cooling modes during equipment maintenance (maintenance repair modes).
Reactor startup and shutdown
RBMK power units are started up with the main circulation pumps in
operation, at a "sliding" pressure and at a separator water level selected by
the operator within a given range. The required cavitation margin at the main
circulation pump intake is ensured by reducing pump delivery using the
throttle-regulating valves installed at the pump discharge. Under these
conditions the cooling water flow rate in all the fuel channels of the core is
continuously monitored and reactor safety is thereby ensured. Initial heating
of the unit is carried out ata "sliding" pressure in the separators, i.e. the
pressure is not constant but increases as the temperature rises.
During the startup and initial heating of the unit'the circulation loop
is fed by the emergency feed pumps. Reactor power during startup and initial
heating is maintained at an average level of 2-3% of nominal capacity. The
thermal power of individual fuel channels during this process can be as much
as 6% of nominal because of the non-uniformity of power density distribution
in the core.
Reactor power ascension and initial heating of the circuit can take
place with one, two or three of the main circulation pumps (capacity
6000-7000 m3/h each) operating on each side of the reactor. At this pump
capacity it is possible to monitor the water flow rate through each fuel
channel and at the same time to ensure an adequate pump cavitation margin. At
a reactor power of 2-3% of nominal, the circuit installations are heated to a
temperature of about 200?C. The circuit is heated at a rate of about
100C per hour, the limiting factor being the thermal stresses in the reactor
metal structures.
At a pressure of 2-4 kgf/cm2, the de-aerators begin to heat up.
A vacuum begins to build up in the condensers of the turbine being
started at a separator pressure of about 15 kgf/cm2. Once the vacuum has
been created, the turbine starts up and begins to build up speed. The
turbogenerator is normally synchronized and connected to the grid when the
pressure in the separators is about 50 kgf/cm2. Further increase in the
parameters up to rated values takes place in parallel with the build-up of
electric load.
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, Figure 2.6.2 gives an example of the evolution of the main reactor
parameters from the time the reactor reaches the minimum power level that can
be monitored until the turbogenerator is synchronized and connected to the
grid.
The main circulation pumps remain in operation during scheduled
shutdown and cooling of RBMK units. Before the onset of shutdown cooling, the
reactor power is run down to the after-heat level and the unit turbogenerators
are disconnected from the grid and shut off. When reactor power is reduced to
the 20% level, the capacity of the main cimulation pumps in service should be
cut to 6000-7000 m3/h. The circuit is cooled down to a temperature of
120-1300C by gradually lowering circuit pressure by discharging steam in a
controlled manner from the separators to the turbine condensers or to the
process condenser. To achieve a greater degree of cooling, a special shutdown
cooling system is employed which consists of pumps and heat exchangers.
The factor limiting the cooling rate, and also the heating rate, is the
thermal stresses in the reactor metal structures. Since during shutdown
cooling the rate of temperature reduction in the circuit is determined
principally by the rate of controlled steam discharge from the separators, it
is not difficult to keep the cooling rate at the prescribed level under these
conditions.
Unit operation at power
During power operation of the unit, reactor safety is ensured by
keeping its critical parameters within the permissible range.
Up to the 500 MW(t) power level, the coolant is circulated through the
reactor by the main circulation pumps operating at 6000-7000 m3/h. At a
power of 500 MW(t), the throttle-regulating valve is opened and the main
circulation pump capacity increases to 8000 m3/h. At power levels above
500 MW(t) up to the rated level, the unit operates at a constant main
circulation pump capacity. When the power level exceeds 60% of rated, no
fewer than three main circulation pumps should be operating on each side of
the reactor. The hydraulic distribution of an RBMK reactor core is such that,
when rated capacity is reached, the throttle-regulating valves are fully open
and the total flow through the reactor is 48 000 m3/h.
Maintenance/repair modes
The main requirement when inspecting or servicing any item of reactor
equipment is that the core must be safely cooled throughout this period.
Also, the reactor design and the organization of maintenance work should be
such as to ensure that all the circuit equipment can be serviced.
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From the standpoint of maintenance work, the primary coolant circuit is
split into four sections: the discharge (delivery-side) section, which
extends from the discharge valves of the main circulation pumps to,the channel
isolating and regulating valves; the fuel channel ducts from the iiolating and
regulating valves to the separators; the separators and downcomers to the
intake valves of the main circulation pumps; and the section between the
Intake and discharge valves which includes the circulation pumps and the
associated fittings.
The maintenance of equipment and pipes located in the section between
the intake and discharge valves of the main circulationpumps does not pose
any difficulties, and in theory can be carried out while the reactor is
operating.
To do this, it is necessary to close the isolating and intake gate
valves on the pipes of the main circulation pump in question; once the coolant
has been drained, the pump itself and the discharge and intake pipe sections
adjacent to it as far as the gate valves are accessible for servicing. In
this instance the coolant is circulated through the reactor by the other main
circulation pumps of the relevant loop.
To repair structural elements of fuel channels, the fuel assembly is
withdrawn from the channel under repair, the isolating and regulating valve at
the channel inlet is closed and the water level in the separators is lowered
to below the level at which the steam-water communication pipe of this channel
is connected to the separator casing. The remaining channels in the core are
cooled either by forced or natural coolant circulation.
To repair the separators, downcomers and intake valves of the main
circulation pumps, the discharge valves of these pumps are closed and the
level in the fuel channels is lowered. To ensure safe cooling of the core
under these conditions, a special maintenance tank is connected to the main
circulation pump pressure header; the channels are fed from this tank and the
steam which forms in them is evacuated to the separators. To allow inspection
and repair of the separators, a system has been installed which draws off
steam from the separators to the process condenser.
During repair work on the equipment of the discharge section, this
section is cut off from the core by closing the isolating and regulating
valves, and the residual heat is removed by water fed into the channels from
the separators. This mode of fuel channel cooling (the bubbling mode of
cooling) was studied on special test units during the reactor design stage.
It was established experimentally that, when the isolating and regulating
valves are closed, safe cooling Of the fuel channels in the bubbling mode is
ensured when the following conditions are met:
The water level in the primary circuit is higher than the levels
at which the steam-water communication pipes connect with the
separator;
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t\\
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The pressure in
the separator is atmospheric;
The after-heat in the fuel assembly is not greater than 25 kW;
The water temperature in the separator is not more than 80-900C
in order to prevent water hammers in the steam-water communication
pipes.
The most complicated repair operation relates to the channel flow
meters and the isolating and regulating valves. To do this work, a technique
is used whereby the water is frozen in the inlet pipes of the fuel channels.
When this technique is employed, the fuel assemblies are cooled in the same
manner as when repairs are being conducted with the isolating and regulating
valves at the fuel channel inlet closed.
The water is frozen in the vertical sections of the inlet pipes by
means of group and single refrigerating chambers attached to these pipes. The
refrigerant is air at a temperature of -1000C which is supplied from the
nitrogen-oxygen station. While the freezing operation is taking place the
Isolating and regulating valves remain closed and the fuel assemblies are
cooled by the bubbling mode. Once ice plugs about 0.5 m high have formed in
the pipes, the isolating and regulating valves and the flow measurement
detectors are accessible for repair. This freezing method has repeatedly been
used successfully at the Leningrad and Chernobyl' nuclear power plants.
2.12.2.Transient modes resulting from equipment failures
Because of the large unit power of RBMK boiling-water, graphite-
mo erated reactors and their extreme importance in energy systems, the control
ad protection system (CPS) of such reactors provides for rapid controlled
power reduction at a prescribed rate to safe levels in the event of the
failure of certain types of equipment. When a signal is transmitted
indicating a fault in the process installations, emergency protection systems
of three kinds (AZ1, AZ2, AZ5) are triggered.
The following algorithm for the operation of emergency protection
syst./ms has been developed for the CPS of existing RBMK-1000 reactors:
AZ1 is triggered when one of the six main circulation pumps shuts
off, the feedwater flow rate decreases and the level in the
separators is reduced. At the AZ1 signal, reactor power is
reduced to the 60% level;
AZ2 is activated in the case of emergency load shedding or the
failure of one of the two operating turbogenerators. At this
signal, the reactor power drops to the 50% level;
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In other accident situations caused by equipment failure the AZ5
emergency protection system is activated and triggers an uncontrolled power
reduction to complete shutdown;
In order to study emergency conditions at RBMK units, a mathematical
model of the plant was developed at the design stage which contains kinetic,
hydrodynamic and heat-exchange equations and a description of the' algorithms
for the operation of the equipment and systems which automatically regulate
NPP parameters. A subsequent comparison of the theoretical results with data
on individual dynamic regimes actually experienced at operating nuclear power
plants indicated that the mathematical model developed provides a satisfactory
description of unit dynamics. Transient regimes mainly associated with
transition to natural circulation of the coolant have been studied on special
mock-up test stands.
Operating experience from units in service has shown that the measures
and systems foreseen guarantee the safety of RBMK reactors in ,all-modes
resulting from equipment failures.
A great deal of research has been done to demonstrate the safety of
reactor operation in the power reduction mode when the AZ5 emergency
protection system is activated since this mode is accompanied by major changes
in the process parameters and, in particular, by a reduction in the water
level in the separators.
The behaviour of the main reactor parameters under transient conditions
due to the activation of the AZ5 protection system is shown in Figure 2.6.3.,
A loss of power plant internal load is ,one of the most severe accident
situations that can occur at the unit. When internal load is lost, the
coolant is circulated through the core at the start of the accident by the
running down main circulation pumps and thereafter by natural circulation.
The transient mode resulting from the loss of the internal load of the unit is
shown in Figure 2.6.4.
This figure shows that, in the initial phase of the process, the
decrease in the water flow rate is somewhat higher than the rate at which the
reactor thermal power decreases; this results in a brief increase in steam
content and reduction in the departure from nucleate boiling (DNB) ratios.
More detailed studies have shown that under such conditions the reduction in
DNB ratios - even in those channels which are under greatest thermal stress -
is insignificant and poses no danger to the reactor, since in the initial
phase of an accident the reactor is safely cooled by the running down main
circulation pumps.
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The running down pumps have a significant effect on coolant circulation
through the reactor only for the first 30-35 seconds of the transient regime.
Thereafter the core is cooled by natural circulation. The reliability and
degree of natural circulation depends to a large extent on a number of factors
such as the primary coolant circuit design, pressure behaviour in the circuit,
the change in feedwater flow rate and temperature and so on.
Experimental research on natural circulation regimes has been conducted
both on heat engineering mock-up rigs of the reactor primary coolant circuit
and directly on operating reactors at the Leningrad and Kursk nuclear power
plants. The experiments at the test rigs established, and those at reactors
confirmed, the safety of cooling the core by natural circulation both in
steady state and in dynamic regimes, given a constant pressure in the
circuit. At the operating reactors, the tests under steady state conditions
were conducted at a power level of 5 and 10% of nominal, while under dynamic
conditions the main circulation pumps were switched off at a power of 25% and
50% of nominal. When the pressure drops as a result, for example, of safety
valves opening and then not closing tightly, the coolant boils, the level in
the separators increases and, as a result, the steam-water mixture is removed
from the circuit. It was established at the test rig that, in the case of
pressure reduction to a certain level, partial removal of the steam-water
mixture and of the water from the circuit does not reduce the "levelling" head
or stop coolant circulation. Overheating of the experimental channel fuel
elements was observed only when the pressure in the separators dropped below
35 kgf/cm2.
To ensure reactor safety follwoing a loss of internal load of the unit
and a sharp drop in pressure, the emergency core cooling system is activated
and feeds water to the fuel channels.
The safety of natural circulation regimes at RBMK units has been
confirmed by accident situations which have occurred under real operating
conditions at nuclear power plants. For example, at one unit of the Kursk NPP
in January 1980 a total loss of station internal load occurred. During the
transient conditions, readings from the thermocouples of the fuel assemblies
and from the flowmeter at the inlet to one of the reactor fuel channels were
recorded. During the entire transient regimes, no increase in the temperature
of the fuel element cans was registered and the flow through the channel
recorded under natural circulation conditions was not less than 20% of the
flow rate at nominal capacity. The normal system for monitoring fuel can
integrity showed no increase in coolant activity when the reactor power was
subsequently increased. Experimental data on natural circulation regimes was
correlated and compared with the results of calculations from the theoretical
programs developed. In view of the good agreement between the results,
theoretical predictions were made which showed that reliable and safe
operation of RBMK-1000 units under natural circulation conditions is possible
at power levels up to 35-40%.
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The reactor loss of feedwater and the separator level protection system
ensures the safe operation of these units at all power levels. When the AZ-5
loss of feedwater emergency protection system is triggered, not only are the
emergency systems activated but the main circulation pumps are disconnected
after a certain time lag. This is done to stop the level in the separator
dropping too much and to prevent cavitation disruption of the main circulation
pumps, i.e. to ensure optimal conditions for effective natural circulation.
As indicated above, the safety of disconnecting the main circulation pumps and
of reactor shutdown cooling by natural circulation has been confirmed by
numerous experiments and by operating experience from nuclear power plants.
Figure 2.6.5 shows the theoretical transient regime following total
instantaneous cut-off of feedwater flow.
The safety of the reactor following accidents in the feedwater supply
system has also been confirmed by operating experience at RBMK units.
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pI rif_t NMI!, w TteC
tb
20
1400'
ISD
200
450
400
FOR OFFICIAL USE ONLY
0 2 Is G. 8 40 42 44 if)
ill Time, h
hg 2.6.2. Evolution of reactor para?eters during startup
I - Water temperature (T) in reactor circulation loops
Pressure (P) in separators
Ther?al power (N) of reactor
3
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41-41P-1
0-70
?0:2?
GG -
- ?
?026 ?
?0,S - - 58
62
-4
Fig
kgf
4,
025
2 . E 3.
40 20 30 4 0
6 0 . 0 D time, s
Rehaviour of reactor parameters following activation of
emergency protection system AZ-5
? Nuclear power
2 - Ther?al power
(N)
)
3 - Circulating water flow rate CO
4 - Change in the level in the separators (pH)
5 - Pressure in separators (P)
245
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k g f
0
?02
?0,
70?.,
62
58
0,25
Fig 2.6.4.
10
20
30
50 GO
Aamomm???????????,
70 80
Behaviour of reactor parameters following loss
of unit internal load.
For legend. see Fig 2.6.3.
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time, s
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7
-
?0,6-
66 0,75
62 0,5
58 125
Fig, 2.6_5.
0 AO 20 30 140 50
60
Transient process following a total
instantaneous cut?off of feedwater.
For legend, see Fig. 2.6.3.
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:ANNEX 3
ELIMINATION OF THE CONSEQUENCES OF THE ACCIDENT AND DECONTAMINATION
3. ELIMINATION OF THE CONSEQUENCES OF THE ACCIDENT AND DECONTAMINATION
3.1. -'PrOgress and prospects for decontamination and startup_of the first,
- second'And"third units
The surfaces of the equipment and ,compartments of the nuclear power
plant were contaminated mainly through the ventilation system which continued
to operate for some time after the accident at the fourth unit and also as a
result of the spread of radioactive dust from the site of the plant. The
highest levels were recorded for separate horizontal sections of the surfaces
in the turbine building (up to 106 0-part/(cm2/min), since it was
contaminated for a prolonged period through the damaged roof.
The y-radiation dose rate in the contaminated compartments of the
first and second units on 20 May 1986 was 10-100 mR/h and in the turbine
building 20-600 mR/h.
The washable nature of the materials (plastic, steel, concrete and
various coverings) and the nature and levels of the surface contamination were
taken into account when choosing the composition of the decontaminating
solutions.
The spraying decontamination method was widely applied in the washing
process making use of washing machines and fire hydrants. Some of the
compartments were washed manually by wiping with a rag soaked in
decontaminating solutions. The steam ejection method was also employed as
well as dry decontamination methods using polymer covers.
The decontamination processes were monitored by direct measurement of
the gamma background from the washable surfaces and by the smear method. As a
result of decontamination, the contamination levels of the surfaces of the
compartments and equipment were on the whole reduced to the norms established
by the Radiation Safety Standard No. 76 and the Basic Health Regulations
No. 72/80:
For service compartments - 2000 0-part/(cm2min);
For semi-service compartments - 8000 ft-part/(cm2min).
After decontamination, the gamma radiation levels dropped by a factor
of 10-15 and the y-radiation dose rate for the compartments of the first and
second units was 2-10 mR/h.
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3.2. Progress and prospects for decontamination of the power plant site
During the accident, radioactive material was scattered over the site
of the plant and also fell on the roof of the turbine building; the roof of
the third unit and on metal support pipes.
The plant site, the walls and the roofs of the buildings" also had
considerable contamination as a result of the fallout of radioactive aerosols
and radioactive dust. However, the overall gamma background at the site
consisting of radiation from the destroyed fourth unit greatly exceeded the
radiation levels from the contaminated site and buildings. 'It should be noted
that the contamination of the site was uneven.
In order to reduce the spread of radioactive contamination in the form
of dust, the site, the roof of the turbine building and the sides of the roads
were treated with rapid polymerizing solutions to reinforce the upper layers
of the soil and to prevent the formation of dust.
In view of the complex nature of the work the nuclear power plant site
was divided into zones for the purposes of decontamination.
The sequence of work carried out for each zone ,was based on the
following criteria:
- The need for staff to work at facilities inside the zones;
- The principle "from dirty to clean" and taking account of.., wind
roses;
- The need for subsequent work involved in startup of the units.
Decontamination in each 'zone was carried out in the following order:
Removal of debris and contaminated equipment from the site;
Decontamination of roofs and external surfaces of the building;
- Removal of a layer of soil, 5-10 cm thick, and transportation of
it in containers to repositories (the solid waste storage vault of
the fifth unit);
- Laying, where necessary, o
clean soil;
concrete slabs or filling in with
- Covering of slabs and non-concreted parts of the site with
film-forming material;
- Restriction of access to the treated site.
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The total number of sites treated ranged from 15 000-35 000 m2 per
24 hours. As a result of these measures the overall gamma background in the
area around the first unit was reduced to 20-30 mR/h. The fact that the
residual background was caused mainly by external sources demonstrates that
the decontamination of the site and buildings was fairly effective. However,
significant improvement in the radiation conditions over the whole site of the
nuclear power plant and particularly in the areas around the third and fourth
units will be possible only after the destroyed reactor has been enclosed.
3.3. Progress and prospects for decontamination of the 30-km zone and its
return to economic activity
The formation of the radioactive trail following a single release of
effluent ends after about a year. After this period there is a significant
redistribution of the radionuclides among the elements of the landscape in
accordance with the characteristics of the relief. The most intensive
redistribution of radioactivity (secondary transfer) occurs during the first
3-4 months after the release, particularly during the course of active
biological and atmospheric processes (growth, development and drying off of
plants, rains and winds). The loosely attached part of the radioactive
substances which have settled on the surface of the soil and vegetation are
subject to considerable redistribution. In coniferous forests such
redistribution ends only after 3-4 years (after complete renewal of the
needles).
For these reasons the radiation conditions within the 30-kilometre zone
will continue to change significantly for 1-2 years particularly in regions
with a high contamination level gradient.
Therefore the measures taken to decontaminate populated areas generally
result only in a temporary improvement in the radiation conditions.
All this leads one to conclude that the evacuated population can only
return to the area after the radiation conditions over the whole territory of
the contaminated zone have stabilized (when releases from the reactor have
ceased, the industrial site has been decontaminated and the radioactivity has
been fixed over the territory where there is an increased contamination
level). Conditions will stabilize most rapidly in regions of the zone with a
low contamination level gradient (for example in the northern and southern
projections of the radioactive path).
In order to decide whether agricultural production can be resumed and
the evacuated population can return, it is necessary to have reliable
information about the concentrations of long-lived radionuclides
(strontium-90 caesium-137) in the soils and the crops cultivated on those
soils. Soil samples have now been taken from all the fields at the collective
and state farms in the region. When these samples have been analysed,
cartograms will be drawn up showing the contamination of the agricultural
lands and indicating the radionuclides.
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Other radionuclides making up the contamination (zirconium-90, '-
niobium-95, ruthenium-103 and 106, cerium-141 and 144, caesium-134,
barium-140, strontium-89) and accounting for more than 90% of the total
activity will not be limiting, factors in the future either because they have
short half-lives or because they are not easily absorbed from the soil by
plants.
In principle the contaminated lands could be reused for agricultural
purposes. General organizational and technical principles for agricultural
management under such conditions have been worked out and numerous
recommendations established for specific aspects. Since the agricultural
conditions of Poless'ye are very specific and the nature of the radioactive
contamination has not yet been studied in detail, specific evaluations can be
made only when the specific data have been obtained. Resumption of
agricultural activity in these areas requires:
(a) The re-organization of agricultural specialization in accordance
with the contamination levels of the lands used; exclusion of
production and produce from directly entering human food; primary
seed production, industrial production and animal fodder
production;
(b) The implementation of special measures aimed at the durable
fixation and consolidation of radionuclides in a form which is
inaccessible to plants for a prolonged period with subsequent
cultivation by applying sorbents (clayey suspension, zeqlites) to
the upper contaminated layer of soil;
(c) The implementation of special decontamination measures involving
the removal of the contaminated surface layer of turf directly by
mechanical means or after consolidation using chemical agents
(latex emulsion SKS-65 gp).
The measures taken to enable the land to be reused for agricultural
purposes will be differentiated according to the time and level of
contamination of the territory.
In the evacuation zone and in the strict control zone, agricultural
harvesting work is being carried out as normal in accordance with the special
measures worked out together with the State Agricultural Programme of the USSR
and Ukranian SSR and the USSR Ministry of Health.
With regard to the surface contamination of vegetation and soils in
1986, the basic special requirements for the organization and the
technological aspects of the work can be summarized as follows:
(a) To reduce to a minimum the mechanical cultivation of soils with
increased dust formation;
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(b) Grain and industrial crops are being harvested by direct combine
harvesting and depending on the actual contamination levels of the
produce are being used (after being stored) for food purposes,
fodder, seed and industrial reprocessing;
(c) A compulsory requirement after the harvesting of perennial grasses
and winter crops is the introduction of lime, mineral fertilizers
and sorbents to increase soil fertility and reduce the entry of
radionuclides into agricultural production.
r
When 'considering the fate Of the.contaminated forests one has to bear
in mind their well-known role as absorbents and accumulators and conservers of
moisture in forest-steppe and steppe regions-.
Research has also shown that in conditions of radioactive
contamination, forests also act as accumulators of radioactive substances,
first in the crown then in the forest litter. The radionuclides fixed in the
litter will for a long time be excluded from the radiation chains.
Therefore at present, the majority of experts believe that the best way
of dealing with the contaminated forests is to increase the fire-prevention
service.
At present, special agrotechnical and decontamination measures, which
are designed to enable the contaminated lands to be reused for economic
purposes, have been developed and are being implemented based on the
evaluations of the contamination conditions of the soil and vegetation cover
in the 30-km zone. These measures include changing the traditional system of
soil cultivation in this region, the use of special dust-suppression
compounds, changing the harvesting and crop processing methods and so on.
The level of radioactive contamination of houses and buildings in the
countryside in the 30-kilometre zone fluctuates within significant limits.
Typical building materials are bricks, wood (boards) both unpainted and
painted, where the condition of the paint varies, slate and roofing iron.
Decontamination was carried out by spraying the surfaces with
decontaminating solution at a flow rate of 10-15 L/m2. Automatic filling
machines were used.
As a result of decontamination, the radiation dose rate from the
buildings dropped to the background levels for this region generally, the
0 contamination did not exceed 1000 0-part./(cm2 min).
After washing the buildings, the radioactive contamination of the earth
along the walls increased by 2-2.5 times and therefore this earth was dug over
or removed with bulldozers and taken away.
The transport vehicles were decontaminated by the spraying and steam
ejection methods using the above solutions.
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ANNEX 4
ESTIMATE OF THE AMOUNT, COMPOSITION AND DYNAMICS OF THE
DISCHARGE OF RADIOACTIVE SUBSTANCES FROM THE
DAMAGED REACTOR
4.1 Amount of radioactive substances discharged from the reactor
The initial information for estimating the amount of these substances
discharged from the damaged reactor consisted in the material obtained from
aerial gamma-ray photography of the region of the Chernobyl' plant and the
national territory carried out by the UNKhV with helicopters of the Air Force
and the USSR State Committee on Hydrometeorology and Environmental Protection
(Gaskomgidromet) as from 1:05 a.m. on 25 June 1986.
In order to determine the amount of radionuclides, the data from the
aerial gamma-ray photography were plotted on a map of the district, isodose
lines were drawn and the areas encompassed by these curves were, calculated.
The results of the estimates of this amount, as at 26 June 1986, are given in
Table 4.1 in absolute and relative values.
The data of this table show that the total radioactivity of the fission
products discharged from the damaged reactor and which had settled on the
ground in a 30-km zone, amounted to 8-14 MCi. An analysis of the findings
obtained showed that at the time the intense discharge of fission products
from the reactor ceased on 6 June 1986, the amount of radionuclides present in
the 30-km zone was approximately 20 MCi. It should be noted in particular
that more than half of this activity is found in a zone with R > 20 mR/h, in
an area consisting all told of 17% contaminated land and including the grounds
of the Chernobyl' plant.
According to an analysis of Goskomgidromet's aerial gamma-ray
photography, beyond the limits of the special zone, the activity of the
radionuclides settling on the ground was 10-30 MCi. It follows from an
analysis of the data that the total activity of the radionuclides released
from the disabled reactor to the environment does not exceed 50 MCi, i.e. it
represents approximately 3-4% of the total activity of the fission products in
the reactor of the fourth unit of the Chernobyl' plant on 6 May 1986.
An independent estimate of the amount of fission products discharged
from the damaged reactor was made by experts of the V.G. Khlopin Radium
Institute. The amount of fuel present in these zones was determined on the
basis of scanning with a collimated detector from a helicopter flying over the
production area at a height of 300 m, analysis of samples taken in the 30-km
zone and use of correlation ratios between the gamma activity of Ce and the
alpha activity of Pu. This value is somewhat higher than the one obtained
from the analysis of the isodose data. The relative values for the
distribution of fission products over the production area and on the roof of
the Chernobyl' power plant building are given in Fig. 4.1.
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4.2. CoTposition of fission products of uranium and other radionuclides
released from the dammed reactor
As the initial information for estimating the composition of the
radionuclides released by the damaged reactor, use was made of radiometric
investigations of aerosol samples and soil specimens carried out by the
V.C. Khlopin Radium 'Institute and the 1.V. Kurchatov Atomic Energy Institute
in the period from 6 to 30 May 1986. These data led to the conclusion that
the composition of the fission products (except for gaseous 1, Te, Cs)
discharged by the damaged reactor is similar to that of the fission products
in the fuel in the reactor itself. This is confirmed, in particular, by the
averaged data from studies of the soil and grass cover in the zone from 1.5 to
30 km from the reactor. These are shown in Table 4.2.
In the 30-km zone, dozens of soil .samples were examined for their
content of transuranium elements with respect to alpha radiation. The
radioactivity of the samples ranged from 2 to 2000 Bq/g and was dependent on
90% 242 Cm. Approximately 10% of the alpha radioactivity was associated with
isotopes of masses 238, 239 and 240. The radioactivity of 238Pu is about
40-70% relative to the sum of the radioactivities of the nuclides 239Pu and
240pu.
The soil sample in a radius of 1.5 km having the relatively largest
concentration of transuranic elements may be considered one found in a
discharge in a south-western direction at the end of a sector with a
contaminated forest. The results of an analysis of this sample, taken on
8 May 1986 on the surface of a road, are shown in Table 4.3. The total alpha
radioactivity of the sample is 1.3 x 104 Bq/g.
The radioactive composition of alpha emitters in the air samples
(filters) and soil samples, according to the data of the 1.V. Kurchatov Atomic
Energy 'Institute are shown in Tables 4.4 and 4.5, and according to the data of
the Khlopin Radium Institute in Table 4.6.
Investigations of the aerosol composition of air samples (with pumping
through a filtering tissue) also confirm the transport via airborne dust of
both volatile and slightly volatile chemical elements without distinct
fractionation, with the exception of iodine, ruthenium and tellurium. The
aerosol samples were taken at a height of 200 m above the damaged reactor and
at a height of 3 in above the earth at 10 fixed points in the production area.
Table 4.7 shows the results of the aerosol measurements at a height of 200 m,
and Table 4.8 at a height of 3 m. Points 3 and 10 of Table 4.8 refer to the
northern direction from the Chernobyl' power plant building and points 8 and 9
to the southern direction. All four points are located on a line running
approximately 150 m more to the east of the damaged reactor (see Fig. 4.1).
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The data of Table 4.7 indicate a sharp drop in specific aerosol
activity after 6 May 1986, which is evidence of the dynamics of fission
product discharge from the damaged reactor with time. Up to 7 May 1986, the
reactor was a source of elevated release of radionuclides, but after 6 May
1986 it ceased being in any degree the determining factor in the formation of
aerosol radioactivity above the production area. This formation was
determined initially by processes of dust formation and secondary wind
transport of radionuclides over the production area as a whole (Table 4.9).
The concentration of radioactive aerosols at a height of 200 in was the
same as that at a height of 3 m (data of 9 and 11 May for 200 in and of 12 May
for 3 m). After 12 May, the concentrations of aerosols at a height of 200 in
become approximately 100 times lower, but at 3 in above the production area
they subsequently showed little change. This can be seen by comparing the
data of Table 4.8 with those of Table 4.9. The latter gives the results of a
determination of aerosol concentrations at a height of 3 m above the
production area, carried out on 22 May 1986 at points located relatively close
to the points referred to in Table 4.8 (see Fig. 4.1).
The compositions of the air and fallout samples showed the presence of
"hot" particles enriched primarily in radionuclides of one type. Fig. 4.2 and
4.3 show the results of measurements of the radionuclide composition of such
particles. As can be seen, there are particles containing practically nothing
but Cs or Ce. There is a tenfold increase in the content of 140Ba over the
theoretical value.
4.3. Dynamics of radionuclide discharge from the damaged reactor
The material used as initial information for analysing the ,dynamics of
the discharge of radionuclides from the damaged reactor was data from
systematic studies of the radionuclide composition of aerosol samples taken
above the fourth unit of the Chernobyl' plant on 26 April 1986. The results
of the investigations are shown in Table 4.10 and in Fig. 4.2.
Analysis of these data led to the conclusion that the release of
radionuclides beyond the limits of the damaged nuclear power plant unit was a
process extended over time, consisting of several stages. The dynamics of the
discharge process are characterized especially prominently by the data of
Table 4.11, whch presents non-dimensional values of fission-product release
(normalized in terms of 1311) in time.
In the first stage, the mechanical discharge of dispersed radioactive
fuel took place as a result of an explosion in the reactor. The composition
of radionuclides at this stage of discharge corresponded approximately to the
composition of fission production in spent fuel, but enriched n volatile
nuclides of iodine, tellurium and caesium.
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In the second stage, from 26 April to 2 May, the capacity for discharge
beyond the limits of the damaged unit decreased owing to the measures
undertaken to terminate the burning of graphite and the filtration of
substances emerging from the core. In a first approximation, the reduction in
the power of discharge in this period can be represented in the form
Q(,) = Q0e-0.5T (4.1)
where Qo is the power of discharge immediately after the explosion (Ci/d);
T is the time after the beginning of the damage (days).
In this period the composition of radionuclides in the discharge was
also similar to their composition in the fuel. At this stage, finely disperse
fuel escaped from the reactor directly with a flow of hot air and with
products from the burning of graphite.
The third stage of release was characterized by a rapid increase in the
power of radionuclide discharge beyond the limits of the reactor unit. In the
initial part of this stage it was primarily the escape of volatile components,
especially iodine, that was observed; subsequently, the composition of the
radionuclides again resembled that of their composition in spent fuel (on
6 May 1986). The power of fission product discharge in the third stage can be
described by the expression
Q (T) = coast. e aT,
where a = (6-8) x 10-2 1/h.
(4.2)
The fact that the discharge was of this nature was apparently caused by
the heating of the fuel in the core to a temperature above 20000C due to
residual heat release. As a result of temperature-dependent migration of
fission product, and also possible carbidization of uranium dioxide, there was
a leak of fission products from the dioxide and their escape either in aerosol
form or in products of the burning of graphite (graphite particles).
The final - fourth - stage, starting on 6 May, is characterized by a
rapid decrease in the escape of fission products from the fuel and virtual
termination of discharge (Table 4.13), which was a consequence of the special
measures taken, by the formation of more refracting fission products as a
result of their interaction with the materials introduced.
Main conclusions:
1. The total release of radioactive substances (not including radioactive
noble gases) was about 50 MCi, which represents 3.5% of the total amount of
radionuclides in the reactor at the time of the accident. These data were
calcualted on 6 May 1986, with allowance for radioactive decay.
2. The composition of radionuclides in the discharge due to the accident
was notable for its elevated content of volatile iodine and tellurium.
3. Generalized quantitative information concerning the variation in power
of discharge with time and composition of the radionuclides released from the
damaged reactor are given in Table 4.13 and 4.14 and in Fig. 4.4.
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Table 4.1 Estimate of tne amount ot radionuclides on the ground in a
30-km zone of the region of the Chernobyl plant as of 26 June 1986
Point
No.
! Zone with R, mR/b I
!'
Area,
KM'
Activity,
.Absolute, MCI
1 Relative, %
1
R>20
870
5-8,7
63,0
2
I0