1986-08-17 – Russian Department of Energy – The Accident At The Chernobyl AES And It’s Consequences

1986-08-17 - Russian Department of Energy - The Accident At The Chernobyl AES And It's Consequences

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State ColTITiittee for Using the Atomic EnPrgy of USSR
HHopManH11, no·,n.roToBJJeHHBR AnJI coBeiuanHR
sHcneproe MAr AT3
(25-29 asryCTa 1986 r. BEHA)
Data rrepared for the
International Atomic Energy Agency
Expert Conference
(25-29 August 1986, Vienna)
-P -U -8 -L -t -C -A -T -l -0 -N
1986 r .
AUGUST 17, 1986
1. Descriptioo of the Chernobyl’ AES with RBMK-1000 Reactors
2. Chronology of the ‘Develq::ment of the Accident
3. Analysis of the Process ·Of Develoμnent of the Accident on a ¥.aat..’1-ierratical
Mcxiel .
4. causes of the Accident
5. Initial Measures to Increase Nuclear PCMer Plant Safety with RBMK Reactors
6. Preventing Developrent of an Accident and Reducing Its Consequences
7. M:mitoring RaC..ioactive Contamination of the Environrrent and the Health of
the Pop.llation
8. Rea:mrerrl.ations for Increasing the Safety of Nuclear Pa.Yer Engineering
• Developrent of 1′.uclear Pc:Me.r Engineering in the USSR

The information presented here is ba.sed. on conclusions of the
Goverrment Carmission on the causes of the accident at the fourth milt of
the Chernobyl’ Nuclear Power Station and was prepared by the following
experts enployed by the USSR State Ccmnission Cotmittee on the.use of Atanic
Aba.gyan, A. A. Mysenkov I A. I.
AsmJlOV f v. G. Pavlovskiy, O. A.
Gus’kova, A. K. Petrov, V. N.
Dernin, v. F. Pikalov, V. K.
Il’in, L. A. Protsenko, A. N.
Izrael’, Yu. A. Ryazantse\•, Ye. P.
Kalugin, A. K. Sivintsev, Yu. V.
Konviz, v. s. SUkhoruchkin, v. K.
Kuz’min, Ia I. Tokarenko, ·v. F.
Kuntsevich, A. D. Khrulev, A. A.
I.egasov, V. A. Shakh, O. Ya.
:Malkin, S • D.
Materials obtained fran the follONing organizations were used in
preparing the infonnation: The Io V. Kurchatov Institute of Atanic Energy,
the Scientific Research am Design Institute of PCMer Equiprent, the
v. G. Khlopin P.adium Institute, the s. Ya. Zhuk “Hydrodesign” Institute, the
All-Union Scientific Research Institute on Nuclear Power Stations, the

Institute of Biophysics, the Institute of Applied Geophysics, the f?tate .
· carmittee on Nuclear Energy, the State Camlittee on hydraneteorology
Ministry of Heal th, the State Ccmni ttee on Nuclear Safety, the Ministry of
Defense, the Main Fire Protection Administration of the Ministry of Internal
Affairs and the USSR Academy of ·Sciences.·

• An accident occurred at the fourth unit of the Chernobyl’ Nuclear Power
Station on April 26, 1986, at 1:23 AM with.damage to the active zone of the
reactor and part of the building in \obich it was located.
The accident occurred just before stowing of the powerplant for
scheduled maintenance during testing of the operating ncxies of one o~ the
turbogenerators. The pc:Mer outp.lt of the reactor suddenly increased
sharply, \tihich led to damage to the reactor and discharging of part of the
radioactive products aCCUltUllated in the active zoi:ie into the atm::>sphere.
The nuclear reaction in the reactor of the fcurth p:Merplant stopped in
the process of the accident. The fire which broke out was extinguished, and
operations -were begun for containing and eliminating the consequences of the
The poμilation was evacuated fran areas irme:iiately adjacent to the
area of the nuclear power plant and fran a zone with a radius of 30 km
around it.
In view of the extrete character of the accident ‘Which occurred at
Chernobyl’, an ~rations grc:up headed by Prine Minister of the u.s.s.R.
N. I. Ryzhkov was organized at the Polithlro of the 0: CPSU (c.entral
Ccmnittee of the Camunist Party of the Soviet Union) for coordinating the
activity of ministries and other g0\7emment departnents in eliminating the
consequences of the accident and rendering aid to the population. A
Govenment Camli.ssian ~ fonred and entrusted with studying the causes of
the accident and carrying out the necessary ercergency and reconstruction
ueasures. The necessary scientific,. technical _and econanic capabilities and
rescurces of the camtry were provided.
Representatives of M1GATE were invited to the USSR and given the .
opportunity to familiarize themselves with the state of affairs at the
Chemobyl’ Nuclear Powerplant and ireasures for overccming the accident.
They infoored the world ccmrunity about their assessrcent of the situation.
The governments of a nmnber of countries, many governmental, social and
private organizations and irrlividual citizens fran various countries of the
world appealed to varirus organizations of the USSR with proposals
concerning participation in overcaning the after-effects of the accident.
SCJre of these propJsa.ls were accepted.
In the thirty years of its develoi;:ment, nuclear power engineering has
.cx:cupied an essential place in \l,10rldwide p::Mer production and, on the whole,
has displayed high levels of sa.fety for man and the envirorment. One cannot
imagine the future of the \l,10rld econany without nuclear power. However, its
further develo?lElt rrust be a~ed by still greater efforts on the part
of science and engineering for ensuring its operatior..al reliability and
‘nle accident at Cllernobyl’ was the result of coincidences of several
events of law probability. ‘!be Soviet Union draws the proper conclusions
fran this accident.

Rejecting nuclear power sources ~ld require a eo~siderable increase
in prcduct.ion and canbustlon of organic fuels. ~s ·’WOlld steadily increase
the risk of h\Dan diseases and the loss.of-water-and forests due to the
continuc:us passage of haIInful chemical substances into the biosphere.
‘nle developtent of the ‘WOrld’s nuclear power resc:urces brings with it,
in addition to gain in the area of the ener~’ supply and-the preservation of
natural resources, dangers of an international character. These danqers
include transfers of radioactivity across bOrders, especially in large-scale –
radiation accidents, the problan of the spread of nuclear -weapons and the
danger of international terrorisn, and the specific danger of nuclear
installations under conditions of war. All this dictates the fun::lanental
·: necessity of deep international cooperation in the field o.f devel~t of
. nuclear power systems and ensuring of their safety.
such are the realities.
The saturation of the m::xlern ‘WOrld with potentially dangerrus
industrial processes, in significantly intensifying the effects of military
operations, places the question of the senselessness and unacceptability of
war under nodern corditians on a new plane.
In a speech on Soviet television on May 14, M. s. Gorbachev stated:
-.ihe indisFUtable ~esson of Chernobyl’ for us lies in the fact that under
conditions of further expansion of the scientific and technical revolution,
questions of the reliability of equiFfiel1t and its safety and questions on
discipline, order anc1 organization take on primary importance. The
strictest requirarents are needed everywhere.
Fu.rthemr:>re, we consider it necessary to IIOV’e toward a serious
deepening of cooper13:tion within the franework of the International ~ency on
Atanic Energy.”
·WI’l’H m1K-1000 RFJ>Cl’ORS
1.1 Design Data
‘!he planned power of the Chernobyl’s Power Station (ChAES) ~ was 60M\7,
and ai January 1, 1986, the pcMer of fa.ir units of the AES was 4000MW. The
third and fourth units belong to the second phase of the ChAES and to the
second generation of these Nuclear Power Stations (AES) •
1.2 Description of the Reactor Installation (RU)
. of the Fourth Unit of the ChAES
The basic design features of REMK reactors are as follows:
1) vertical channels with the fuel and the heat-transfer agent, ‘Which
pennit local reloading of fuel with a ~rking reactor;
2) fuel in the fonn of b.mdles of cylindric fuel elanents of uranium
dioxide in zirconium shell tubes;
3) a graphite m:xlerator between channels;
4) a low-boiling heat-transfer nedium in the forced circulation
rrcirculation ncde (KMP’l’s) with direct feeding of steam to the turbine.
‘lbese design decisions in canbination condition all the basic features
of the reactor and the AES, both· advantages and shortcanings. The
advantages include: the absence of reactor vessels, which are awkward to
produce on the powerplant maximJm capacity and on the production base; the
absence of a catplex and expensive steam generator; the possibility of
tinuous reloading of fuel and a good neutron balance; a flexible fuel
cycle, ~ch is easily adapted to variations in the fuel market conditions;
the :possibility of nuclear superheating of the steam: high themodynamic
reliability of the thel::mal equiprent and viability of the reactor due to the
controlling of the flow rate for each channel separately, JI’Oni toring of the
integrity of the channels, m::>ni toring of the parameters and radio activity
of the heat-transfer m:dium of each channel and replacenent of damaged·
channels while running. The short.canings include: the po~sibili ty of the
developnent of a :positive void coefficient of reactivity due to the phase.
change in the heat-transfer agent which deteI:mines the transient neutronic
behavior; high sensitivity of the neutron field to reactivity disturbances
of different kinds, necessitating a canplex control systarl for stabilizing
the distribution of the release of energy in the active zone: ccrrplexity of
.the inlet-outlet piping system for the heat-transfer agent of each channel;
a large anount of theJ:IPal. energy accunulated in the metal structures,· fuel
elerrents and graphite block structure of the reactor; slightly radioactive .
steam in the turbine.
The RBMK-1000 reactor with a pc::7Wer of 3200 MW (thennal) (Fig. 1) is
equipped with two identical cooling loops: 840 parallel vertical channels
with heat-releasing assemblies (TVS) are connected to each loopo.
A cooling loop has four main parallel circulation pol’C’pS (three working
pmps feeding 7000 t/h of water each with a head of about 1.5 MPa, and one
back-up pmtp).
‘nle water in the channels is heated to boiling and partially
evaporateso ‘!’he water-steam mixture with an average steam content of 14% by
mass is bled through the top part of the channel and a water-steam line into
two horizontal gravity separators. The my steam (with a nDisture. content
less than 0.1%) separated in them passes fran each separator at a pressure

of 7 MPa in two steam lines into two turbines with a JX7Wer of 500 M-7
(electrical) each (all eight steam lines of the foor separators are jointed
by a ocmron •ring”), and the water, after mixing with steam condensate:,· is
fed by 12 down pipes into the intake collector of the main cooling J?.JITPS.
Condensate of the steam exhausted frcm the turbines is returned by feed
water ?JlIPS through separators into the top part of the aa.ID pipes, creating
un:ierheating of the.water to the saturation tanperature at the main cooling
p.l[ilp inlet.
‘ll’le reactor as a whole is made up of a set of vertical channels with
fuel and the heat-transfer nedium l::uilt into cylindric apertures of graphite
colmons, and top and bottan protective plates. A light cylindric housing
. (casing) encloses the space of the graphite block structure.
The blcx::k structure consist of graphite b°locks with a square cross
section with cylindric apertures along the axis assanbled into columns. The
.block structure rests on the OOttan plate, which transmits the weight of the
reactor to a concrete shaft.
Al::out 5% of the reactor ~r is released in the graphite fran slowing
· down of neutrons and absorption of gamna quanta. For reducing the themal
resistance and preventing graphite oxidation, the block structure is filled
with a slowly circulating mixture of helium and nitrogen, which serves at
the sane time for nonitoring the integrity of the channels by m=asuring the
humidity and tanperature of the gas.
There are spaces under the bottan and over the top plates for placing
~t carrier pipes on roo.tes fran the separator dnmls (BS) and distributing
collectors to each channel.
A rooot – a loading.and unloading machine (RZM) – after renoval of the
appropriate section of the plating and after being noved·to the coordinates
of the charmel links with its head, balances its pressure with the pressure
of the channel, unseals the channel, raroves the blrned-out (fuel elemants
(TVS) and replaces them with a fresh one, seals the channel, \ll’lCOOples
itself and transports the irradiated ‘IVS to a holding tank. ‘While the RZM
is connected to the cavity of the channel (TK), a small fl<:7« of i;:m-e water passes fran it through a thenrohydraulic seal into the TK, creating a "barrier" to the penetration of the RZM by hot, radioactive water fran the '!J(. The system for control and protection (SUZ) of the reactor is based on novarent of 211 solid absorber rods in specially isolated channels cooled with water of an independent duct. The system provides: autanatic adjustment to a specified ~ level; a rapid reduction of the power level adjustnent to by OOth rods of autanatic regulators (AR) and rods of manual regulators (RR) according to malfunction signals fran the basic equipnent1 energence interruption of the chain reaction by emergency protection (AZ) rods accordlllg to signals of dangeroos deviation.S of the paraneters of the unit or malfunctions of the equipnent1 cc:up:nsation for reactivity variations in heating up.and emergence at power; regulation of the distrib.ltiai of the release of energy over the action zone. RBMK reactors are equipped with a large number of independent control systans, which are being uoved into the active zone at a rate of 0. 4 ml s in functioning of the AZ. 'nle low rate of IIDV~t of the control ~stans is CCJnFEnSated for by the large nmober of systems. • • 5 The SUZ includes subsystems for lcx::al autanatic control· (LAR) and local srergency protection (I.AZ). Both operate according to signals of ionization chaui::>ers inside the reactor. The I.AR autana:tically stabilizes the
fundamental haJ:m::nics of radial-azimuthal distril::xltion of the release of
energy, while the I.AZ provides errergency protection of the reactor against
~ceeding the specified p::1W’9r of channel cartridges in reactor individual
areas. Shortened absorber rods (USP) intrcxluced into the zone fran the
l:::ottan (24 reds) are included for controlling the ~ fields along ·the
height of the reactor •
The RBMK-1000 reactor includes the following. basic JYDnitoring and
control systans in addition to the SUZ:
l) a system for physical JYDnitoring of the field of the ·release of
energy along the radius (JYDre than 100 channels.) and the height (12
channels) by neans of direct charging pickups;
2) a start-up nonitoring system (neutron flux nonitors, start-up
fission chambers);
3) a systan for m:mitoring the water fla.; rate along each channel with
ball fl~ters;
4) a systsn for m:>nitoring the integrity (KOO) of the fuel elarents
based on m=asuring the .short-tine activity of volatile fission produG:ts in
water-steam lines (PVK) at the ouilet·fran each channel; the activity is
detected sequentially in each channel in appropriate optimJm energy ranges
(”windows”) with a photamll ti plier, which is m::wed fran one PVK to another
by a special carriage;
5) a system for nonitoring the integrity of the channels (Icr’sTK) by
nea.suring the humidity and the te:nperature of.the gas flowing in the
All the data pass to a canputer c The information is given out to the
operators in the foi:rn of deviation signals, indications (on call) and data
of recorders.
UJ:he RBMK-1000 power units operate primarily in a base-load no:ie (at
constant power output).

In view of the great power of the unit, a full autanatic shut-down of
• the reactor occurs only if indicators of the power· 1evel, pressure or water .
level in the separator pass. beyond acceptable limits, in a case of a general.
cut-off of electric current, disconnection of two turl:cgenerators or two
mall1 cooling p..imps at once, a drop in the feedwater flow rate by a factor of
nore than 2, or full cross-sectioned rupture of the main outlet pipeof
cooling pJIIpS with a diarceter of 900 nrn. In other cases of equipnent
failures, only an autanatic controlled reduction in power (to a level·
corresponding to the po.Yer of the equiμrent which has remained in· operation)
is envisaged.
1.3. Basic Physical Characteristics of the Reactor
The RBMK-1000 nuclear :power reactor is a heterogeneous the:anal channel
.reactor, in which uranium dioxide weakly enriched in regard to uranium-235
.is used as fuel, graphite is used as.rroderator and boiling light water is
used as the heat-transfer m=dium. The reactor has the following basic
Thermal power
Fuel enriclment
Uranium mass in a cartridge
Number I diaireter of fuel
elerrents in ‘IVS
Depth of fuel burnup
“Coefficient of non-unifo:oni.ty of
release of energy along the
Coefficient of non-unifoIInity of
release of energy along the
3200 MW
114.7 kg
18/ 13. 6 IIlll
20 MW day/kg
Calculated maximum· power of
Isotopic canposition of
unloaded fuel: ·
Void reactivity coefficient
at a working point
Fast power reactivity coefficient
at a working point
Coefficient of expantion fuel
temperature coefficient
Coefficient of expantion graphite
temperature coefficient
Minirm.Jm “weight” of rods of suz, ~
Effectiveness of r6ds of RR, AK
Effect of replacarent (on the average)
of the burnup ‘IVS with fresh
3,250 kW
4.5 kg/t
2o4 kg/t
2.6 kg/t
1.8 kg/t o.s kg/t
2.0 x 10- ‘°·/ vol.% steam
-o.s -to x 10 /Mil
-S o -1.2 x 10 I c
An i.Irp:>rtant physical characteristic fran the point of view of control
and safety of the reactor is a, value called the operating reactivity margin.
The o:perating reactivity margin neans the specific number of suz_ rods ·
plunged into the active zone which are in a region of high differential
efficiency., It is deteDnined by recalculation for fully sul:nerged SUZ rods.
The value of the reactivity margin for RBMK-1000 reactors is generally
accepted as 30 RR rods~ In this case, the rate of intrcx:luction of a
negative reactivity in functioning of the AZ anounts _to l “f’/s (“~” is the
prqx:>rtion of delayed neutrons), which is sufficient far ccrcpensatian for
positive reactivity effects.
‘Ihe character of the dependence of the effective breeding coefficient
n the density of the heat-transfer n:aliurn in R8MK reactors is .deteIJni.ned to
a great degree ·b’j the presence of absorbers of different kinds in the active·
zone. In initial charging of the AZ, which includes abalt 240 boroncx:
mtaining additional absorbers (DP), dehydration results in a negative·
reactivity effect.
At the same tine, a small increase in the steam content at naninal
power with a reactivity margin of 30 rods results in an increase in
reacti. vi. ty ( =2.0 x 10- a+/ vol.% steam).
For a boiling water~aphite reactor, the basic paraneters which define
its ability to properly operate alli safety in the regard to thennal
equipnent are: . the terrperature of the fuel elarents, the margin before the
a crisis.of heat transferoccurs, and the graphite tarq;:ierature.
A set of canputer codes which makes it possible to conduct operating
:calculations on station carplters for ensuring plant reliability of thenna.l
equipnent of the powerplant in a node of continuous reloading of fuel at any
position of the cut-off and control valves at the inlet to each channel has
been developed for RBMK reactors. Thus the possibility of detennining the
·physical paraneters of the reactor at variable frequency of the adjustnent
of channel flow rates and different control criteria (based on eighter
outlet steam quality or on the margin of the critical power) and also as a
function of the throttling of the active zone is provided.
Far defining the fields of the release of energy over the active zone
of a reactor, indications of the physical m:>ni toring system, based on
nea.surarents of the neutron flC11.• along the radius and height of the active
zone takeri·inside the reactor, are used. In addition to.indications .of the
pbysical m:mitoring system, data characterizing the carposition of the
active zone and the energy generation of each TK, the arrangarent of· the
regulating rods, the di.strihltion.of water flow rates along channels of the
active zone and readings of gages of the pressure and ~rature of the
heat-transfer iredium re also entered into the station c~ter. As a result
of calculations by the PRIZMA program performad periodically by the
ccmputer, the operator receives infonnation on a digital printing device in
the fom of a cartogram of the active zone, which indicates the type of
loading of the active zone, the arrangarent of regulation rcxis, the network
of the arrangarent of pickups inside the reactor, and the distril:ution of
por.r.ier levels, water_flow rates, reserves up to critical powers and reserves
up to the naximlin acceptable thermal loads on the fuel elements in regard to
each fuel channel of the reactor. The station c:arp.Iter also carpltes the
averall the:cmal i;x:rwer of the reactor, the distribution of flow rates of the
steam-water mixture arrong the separators, the integral generation. of power,
the steam content at the outlet fran each TK and other pararreters necessary
for nonitoring and controlling the installation.

‘!he e>..~ience of operation of active RBMK reactors indicates that with
the means for m:mitoring and control available on these reactors, maintain-
~temperature coooitions of the.fuel and the graphite and reserves before
a crisis of convective heat transfer at an acceptable level causes no
1.4. Safety Assurance Systsns (Figso 2 and 3)
1.4.1. Protective Safety Systems
The syst.sn for emergency cooling of the reactor (S1£>R) is a protective
safety system and is intended for providing el..imi.nation of the residual
release of heat by prcmpt. feeding of the required arrount of water into
reactor channels in accidents acccrnpanied by disruption of cooling of the
active zone •
Such accidents include: ruptures of large-diameter KMP’I’s pipelines,
ruptures of steam lin~s, and ruptures of feedwater pipelines.
The system for protection against an excess of pressure in the main
heat carrier duct is intended for providing an acceptable pressure level in
the duct due to rerroval of steam into a perforated sprayer tank for its
The system for protection of the reactor space (RP) is intended for
ensuring that an acceptable pressure is not exceeded in the RP in an
erergency situation with rupture of one operating channel due to rem:wal of
the steam-gas mixture fran the RP into the screen of steam-gas discharges of
the sprayer tank and then into the sprayer tank with simultaneais
extinguishing of the chain reaction with the AZ facilities. The SK>R and
the system for cooling the reactor space can be used for introducing the·
‘ I appropriate neutron absorbers (salts of boron and He).
1.4.2 localizing Safety Systans
‘nle system for localization of accidents (SIA) realized on the fourth
unit of the ChAF.S is intencled for localizing radioactive discharges in
accidents with unsealing of any pipelines of the react.Or cooling duct except
the PVK pipelines, the top tracts of the operating channels and that part of
the down pipes which is located in the separator drum canpartment, and
pipelines for st.earn-gas discharges fran the RPc
The main ~t of the localization system is a system of airtight
canpartments, including the following canpartments of the reactor division:
– tightly packed cells arranged syrnnetrically in relation to the
reactor axis and designed for an excess pressure of 0.45 MPa:
– canpartments of separator group collectors (RGK) and bottan water
lines (NVK); these canpartments do not pennit an increase in excess pressure
above Oo08 MPa according to the conditions of strength of canponents of the
reactor structure and are designed for this value.
carpart:nents of ·tightly packed cells arxi the steam distributor corridor
are connected to the water· space of the perforated sprayer condensation
device by steam outlet channels.
‘nle cut-off and sealing armature system is intended for providing
airtightness of the zone of localization of accidents by cutting off
ccmnunicating lines connecting the sealed and unsealed canpartmen~o •
The rubbling condensation device is intended for condensation of steam
– in the precess of an accident with unsealing of the reactor contour;
– in functioning of the main safety valves (GPK);
– in leaks through the GPK in a nolJilal. operating m:ide~
1.4.3. Security Safety Systans
The AES Pcwer SUwly
Electric p::r.r.ier users at an AES are divided into three groups, depending
on the requirerrents placed on the reliability of the power supply:
1) users who cannot i;:ermit interruption of the feed for fractions of a
second up to a f~ seconds under any conditions, including conditions of a
total disappearance of alternating current voltage fran “”=>rking and back-up
transfonners for system needs, and who require the obligatory presence of a
IX’Wer supply after functioning of the reactor AZ;
2) users who can accept a :p:::Mer interruption of tens of seconds up to
tens of minutes under the sane conditions and require the obligatory
presence of a pciwer supply after functioning of the reactor AZ7
3) users who do not require the presence of a :i;:iower supply in conditions
of a disappearance of voltage fran ‘WOrking and back-up transfonners
for systan needs and in a noDna.l m:rlel of operation ·of the unit can peDilit
interruption of the supply for the t.i.ne of transfer fran a workin:J to a
• back-up transfomier for system needs.
1.4.4. Controlling Safety Systems
Controlling safety systems are intended for autar\atic engagemant of · ·
devices of protective, localizing and security safety systems and for
:aonitoring of their operationo
L4 .5. The Radiation Monitoring System
The AES radiation rconitoring system is a carq;:ionent (subsystan) of the
AES autarated control system and is intended for collection, processing and
display of information concerning the radiation situation in canpart:ments of
the AES and in the external environment, the condition of operating
facilities and ducts, and irradiation doses to personnel in accordance with
active noIJnS and legislation.
__ !

1.4 .6. AES Control Points
Control of the AE.S is carried an at ~- levels: station and . plant.
All the control systans ~ich ensure safety of the AES are located at
the plant level.
1.5. Description of the Area of the Chemobyl’ ·AES
and the Areas in Which It is Located
1.5.1. Description of the Regien
The Chernobyl’ AES is located in the ea.stem part of a large region
kno,..n as the Belorussian-UJcrainian Alluvial Plain, on the banks of the
Pripyati River, which flc:Ms in the Dnepr. This _region is characterized by a
.threela tively flat relief with ver:y slight surface slopes in the direction of river and its.tributaries.
The total length of the Pripyati up to its flow into the Dnepr is
748 km; the area of the drainage basin at the AES site is 106 thousand km ,
and the width is 200-300 m. The average flow speed is 0.4-0.S m/s, and the
average water flow rate over many years is 400 ~ /s.
The water-bearing level, which is used for danestic and drinking water
needs of the region in question, li;es at a depth of 10-15 m in relation to
the current depth of the Pripyati and is separated fran (\la.ternary deposits
by clay marls ~ch are relatively impenneable to water •

·._ .. ~.
‘l.’he region of the Belorussian-Ukrainian Alluvial Plain as a wh::>le is
characterized by a low population density (before the beginning.of
construction of the Chernobyl”· AES,. the. average i;:opulation density in the
region in question was approximately 70 people per km ) •
At the beginning of 1986, the total pc1?.llation in a 30-kilareter zone
ara.ud the AES am::unted to about 100 thoosand people, of whan 49 thousand
lived in the city of Pripyati, located west of the three-kilareter sanitaryprotection
zone of the AES, while 12. 5 thousand lived in the regional
center, the city of Chernobyl’, located 15 km to the southeast of the AES.
1.5.2. Description of the AE.S Areas and Its Structures
‘ll1e first phase of the Chernobyl’ .AF.s, CXllifXJsed of tw:> power units with
RBMK-1000 reactors, was milt in the perio::l of 1970-1977, an:9. construction

of two power uni ts of a second phase was ccmpleted at the sane site by the •
. end of 1983.
Construction of another two power units with reactors of the sane kind
(the third phase of the AF.S) was begun 1.5 Jan southeast of this site in
To the southeast of the AES site, right in the valley of the Pripyati
River, a water cooling pond was tuilt with an area of 22 km
1 the pond
provides cooling of tllrbine condensers and other heat exchangers of the
first four p:iwer units. The no:cmal retaining level of water in the cooling
porXi was adcpted as 3.5 m below the grading mark of the AES siteo

~ high~city cooling towers (a hydraulic load of 100 thousartd·m /h
each), which can operate., parallel with the cooling pond, are being l:uilt as
part of the third phase of the AF.S.
To the west and.north of the site of the first and Second phases of the
AES is the area of the construction base and the supply departnent.
1.5.3. Data on the Number of Personnel at the AES
Site During the Accident
There were 176 duty operating personnel and, also, other workers of
various shops and repair services at the site of the first and second phases
of the Chernobyl’ AES on the night of April 25 and 26, 1986.
In addition, 268 construction workers and assanblers were working on
the night shift at the site of the third phase of the AES.
1.5.4. Infonnation About the F.q\litrrent at the Site Which Operated
Together With the Damaged Reactor and About the F.quiprent
Used in the Procesf:.i of the Overcaning the Accident ·
Construction of the Chernobyl’ AES is carried out in phases, which each
consist of two pc::Mer units and have special water purification systar.s
camon to the ‘bJo units and have special water purification systems cc:rmon
to the two units and auxiliary structures and the industrial site \rfuich
. include:
– storage for liquid and solid radioactive-wastes:
– open c:listrirutor devices:
– gas equipnent·:
back-up diesel generator power plants:
– hydraulic engineering and other structures.
The storage for liquid radioactive wastes,_ built as part of the second
phase of the AFS, is interrled for collection and tarp:macy storage of liquid
radioactive wastes arriving in operation of the third and foorth units and
for collection of water fran operational flushing and its recovecy for
reprocessing. Liquid radioactive wastes pass fran the main housing by
pipelines laid on the bottan level of a scaffold, while the Solid
radioactive wastes ccne to the storage by the top oorridor of the scaffold
by electric trucks.
A nitrogen-oxygen station is intended for satisfying the needs of the
third and fourth units of the AESe
The gas equipnent is made up of canpressor, electrolysis, helimn and
argon tank equipnent in~ for providing the third and foorth units of
the AFS with canpressed air, hydrogen, helium and argon. Receivers for
storing nitrogen and hydrogen are lcx:ated in open areas.
A back-up diesel power plant (RDFS) is an independent emargency _source
of electric power for systems inp::>rtant to the .safety of each unit. Three
diesel generators with a unit power of 5.5 Mi were installed on each RDF.s of
the third and foorth units. Intelltlediate and base diesel fuel depOts, ~
transfers of fuel, and errergency fuel ·and oil drainage tanks are included .
• for ensuring operation ·of the RDFS. ·
The source of the technical water supply for the third.and fourth Units
is the ccx:.>ling pond.
‘nle water of the circulation ?JI1TP house, which is unified for the third
and foorth units, is fed into a delivecy tank, fran which it passes by
gravity fla.iv into the turbine con:iensers.
Separate water.works of the third and fourth units are included for
supplying technical water to :Unportant users ‘Who require an uninterrupted
water supply. A back-up power supply fran diesel generators is available
for these water ‘WOrks.
All four power units of the first and second phases and auxiliary
systems and industrial area facilities involved with their nonna.1 operation
were ‘Werking on April 25, 1986.
The Chernobyl’ Powerplant No. 4 was p.lt into operation in December,
1983. ‘B’j the time of stopping of the plant for a m=dium repair, which was
planned for April 25, 1986, the active· zone contained 1659 TVS with an.
average blrnup of 10.3 MW day/kg, 1 DP and 1 unloaded channel. The main
pa.rt of the TVS {75%} were cartridges of the first loading with a bumup of
12-15 MW day/kg.
Tests of turbogenerator No. 8 in a runout rn:xie with the auxiliary
oonsurnption load only internal needs were planned just before stowing. The
. purpose of these tests was to experim=ntally verify the t:0ssibilities for
·Using mechanical inertia energy of the rotor of a turbogenerator disc,onnecte::
l fran steam suwly, in order to generate electricity for aUX.iliary
IIOtors ‘«hat may be required if the turbogenerator is disconnected fran an
electric grid. This node is used in one of the subsystems of the high-speed
·:.::system for arergency cooling of the reactor (SADR). With the proper order
of perfoIInance of ·the tests and additional safety xreasures, the perfo:r:mance
of tests of this kind on a ‘NOrking AES was not prohibited.
Buch tests had already been perfonned previously at this station. It
was established at that tine that the voltage on the generator busses drops
Dllch before, the rcechanical (inertia) energy of the rotor in running downe
In the tests scheduled for April 25, 1986, the use of a special systen to
control regulator of the magnetic field of the generator, which was to have
eliminated this shortcx:ming, was plarmed. However, the “Working Program of
Tests for ‘l’urbogenerator No. 8 of the Chernobyl’ AF..S” in accordance with

which the tests were to have been conducted was not prepared and approved in
the prq:>er way.
‘!he quality of the program proved low; the section on safety measures
included in it was cc:mposed ?Jrely as a matter of form. (It ?=>inted a;it
only that in the process of tests, all switching is done With the
authorization of the station shift director; in case of develoμrent of an
emargency situation, all pe.rsormel must act in accordance with local·
instructions: and just before the beginning of the tests, the test leader –
an electrical engineer, who is not a specialist on reactor installations –
briefs the watch on duty.) In addition to the fact that the pro;rams
.. essentially included no additional safety measures, it prescribed
.··disengaging the system for em:rgency cooling of the reactor. This ~t
that throughout the period of the tests_, i.e., about 4 hours, the safety of
;the reactor a:pp=ars to have been lowered significantly.
On the strength of the fact that the proper attention was not devoted
to the safety of these tests, the personnel were not ready for them and did
not know about the ?=>ssible dangers. In addition, as· one will be able to
see fran what follows, personnel deviated fran carrying out the program,
thereby creating the conditions for develoi;:rnent of an emergency situation.
‘llle personnel started to reduce the power ou’tplt of the reactor, Wich
had been operating . at nai\inal parameters,. at 1: 00 AM on April 25, and at
1:05 PM turbogenerator No. 7 (‘l’G No. 7) was disconnected fran the grid at a
reactor thermal a.it?Jt of 1600 Ki. The electric~ supply for the
auxiliaries (4 main cooling p.llI’pS, 2 feed waterp.mips) was transferred to
the busses of turbogenerator No. 8. –
‘!be 51-DRwas disengaged fran the Ire main coding p.mips, one fran each side ~e engaged in addition
to the six pumps. which had been operating, so that after the end of the
experiment, in which foor pmps were to operate to sui;p:>rt the runbut nDde ••
of operation, fwr pmps ‘Walld.remabl in the forced circulation loop (KTPI’)
, reliable cooling ·of the active zone.- .
Since the :ceactor power and, consequently, the hydraulic resistance of
the active zone and the lrtant to make sure that the mathematical
m:xiel of the power unit accurately describes the behavior of the reactor
and the other equi:i;,itent and systans under. just those conditions making up
the sitllation just before
the breakdam. As already nentioned in the previous section, the reactor
was operating in an unstable manner after”l:OO AM on .April 26, 1986, and
the operators were introducing “disturbances” into the. control object
practically continuously for stabilizing its paramaters. This made it
possible to carpare actual data recorded with adequate reliability by
recording devices to data optained in nurrerical sim.llation for quite a
large time inteJ:val under various effects on the reactor installation. The
canparison results proved. quite satisfactory, which attests to the adequacy
of the mathematical m:Xie1: and the real object.
In order to present the effect of prehistory on the character of
develqnent of the accident m::ire clearly, we shall analyze the calculation
data beginning fran 1:19:00 AM, i.e., 4 minutes before the beginning of the
test with rundown of the TB (Fig. 4o). This ncirent is convenient in that
the operator began one of the operations for repleni.shrcent of the separator
drums (the second since 1:00), which introduced strong disturbances into
the regulation object. At this m:m:::mt, the DREG program recorded the
positions of rods of all three AR: i.e., the initial conditions for the
calculation were clearly recorded.
‘!he operator began replenisl:’lm;mt of the separator drums to avoid
allowing a dip in the water level in them. He succeeded in maintaining the
level in 30 s, having· increased the flow rate of feedwater by a factor of
nore than 3. The operator apparently decided not only to maintain· the
water level but to raise it. Therefore, he continued increasing the water

• flow rate, and it exceeded the original flow rate. by a factor of 4 in just
al::olt a minute.
As soon as colder water fran the separating drums reached the active
zone, steam generation decreased noticeably, causing a decrease in the
volunetric steam content, which resulted in m:JVatent of ali the AR rods
upwcu:d. In about 30 s they anerged at the top ends, and the operator was
forced to “help” than with manual control rOO.s, thereby reducing the
operating reactance reserve. (‘!bis operation was not recorded in the
operation log, but it wculd have been impossible to maintain power at a
level of 200 ~ withoot it.) The operator, having rrcved the manual rods
up, achieved recanpensation, and one of the groups of AR rcxis was lowered
by 1.8 m.
The decrease in steam generation led to a small pressure decrease •
. · After about a minute, at 1: 19: 58, a high-speed reduction device (BRU-K) ,
throogh which steam surpluses were released into the condenser, was closed.
‘!bis praroted sane decrease in the rate at which the pressure was dropping.
HO’it1eVer, the pressure continued to drop slowly up to· the beginning of the
test. It changed by rrcre than 0.5 MPa during this period.·
A printout of ·the actual fields of releases of energy and the
positions of all the regulation rods was obtained on the “Skala” STsK at
1 :22 :30. An attempt has been ma.de at “tying together” the calculated and
recoJ:ded neutron fields by just this m::ment.
The overall characteristics of the neutron field at this m:::ment were
as fol.lows: it was practically arched in a.radjaJaz:imutha.l direction and
doo.ble-peaked, on the average, iri regard to height, with a higher release
of energy in the top sect.ion of the-active zone. Such a field distribution
quite natural for the situation of the reactor: a depleted active zone,
alm:>st all the regulation rods up, a volunetric steam content significantly
higher in the top part of the active zone than at the bottc:in, contamination
with Xe higher in the central parts of the reactor than in the peripheral
(:,- ~
The reactance reserve anounted to a total of sac rods at 1:22:30.
:nll.s value Wc3.S at lease two tine lower than the minim.Im acceptable reserve
. established by technical operating regulations. The reactor was in an
unusual, nonregulation condition, and for.evaluating the subsequent
develOflTEilt of events, it was extremely inp:>rtant to deteJ:mine the
differential efficiency of rods for regulation and arergency protection in
real neutron fields and the fission characteristics of the active zone.
Nurrerical analysis indicated high sensitivity of the error in detennining
the efficiency of the regulation rods to the error in reconstruction of the
vertical field of releases of energy. If one takes into account in
addition that at such low power levels (about 6-7%), the relative field
xreasuratent error is substantially higher than under naninal conditions,
the need for analyzing an extrsrely large number of calculation versions to
ascertain the reliability or inaccuracy of sare version beCCl’!es clear ..
‘!be reactor pa.raneters were closest to stable for the time pericd in
question by 1: 23, and the tests began. A minute before this, the operator
sharply reduced the fee:iwater·flow rate, which occasioned an increase lli
the water ~ature at the inlet to the reactor with a delay equal to the
~ of passage of the h~t-transf er ·medium fran the separator drums to the
At 1:23:04 the operator closed the. SRI< of rm No. 8 and began rundown of the turbogenerator. Due to the decrease in the flow rate of steam fran the separator drums, its pressure began to increase slightly (at a rate of 6 ~a/s, on the average). The total water flow rate throogh the reactor began to drop due to the fact that foo.r of the eight GTsN were working off , the turbogenerator which was "nmning da.om. n The increase in the steam pressure, on the one hand, and the decrease in the water flow rate through the reactor and also in the feedwater suwly to the separator drums, on the other, are canpeting factors which deteJ::rnine :the volumetric steam cx:mtent and, consequently, the p:TWer of the reactor. It should be anphasized in particular that in the condition at which the reactor arrived., a small change in the p::1W'e!'. results in a situation where the volurretric steam content, which directly influences reactance, increase many tines :rcore sharply than at nani.nal p:rwer. The canpetition of these factors led in the final analysis.to a ~r increase. Just this situation could be the cause for pressing button AZ-5. Pushbutton AZ-5 was pressed at l: 23: 40. Insertion of energency protection rods began. By this tine, the AR rods, in partially ccmpensating for the previous increase in~, were already located in the OOttan part of the active zone, while the "WOrk of personnel with an unacceptably lCM operating reactance reserve resulted in a situation where .ractically all the other absorber rods were located in the top seqtion of the active zone. 30 Under the conditions which. had been created, the disrupt.ions peDnitted by the personnel resulted in -a significant decrease in. the efficiency of the erergency protection. The total positive reactance developing in the active zone began to increase. After 3 s the power exceeded 530 M.v, and the runaway pericd cane to be IlUlch less than 20 s. ~e positive steam effect of reactance prCltDted deterioration of the situation. Only the Dowler effect partially a:xnpensated for the reactance introduced at this tine. The continuing decrease in the water flow rate through the operating channels of the reactor under conditions of an increase in power led to intense steam fonnation and then to a crises of convective heat transfer, heating up of tlie fuel,.its disintegration, rapid boiling of the heat.transfer agent, into which particles of disintegrated fuel were falling, a sharp increase in pressure in the operating channels, rupture of the channels and a thennal explosion, which destroyed the reactor and part of. the structural canponents of the building and led to the release of aci:ive fission products into the envirorment. Disintegration of the fuel was simulated in the mathanatical m:Xiel by a sharp increase in the effective heat-transfer surface area, where the specific release of energy in the fuel exceeded 300 cal/g. At just this tine, the pressure in the active zone increased to the extent that a shaJ:p decrease in the water flow rate fran the GTsN occurred (the chec:k valves closed). This can be seen clearly both fran results obtained on the mathematical m::>del and fran neasurement results recorded by the om;
prcqram. Rupture of the operating channels alone led to partial

reconstruction of the flow rates frcm·the GI’sN, althoogh water passed fran
the:n into the reactor
space as well as into the SUIViving channels.
The steam fonna.t~on and the sharp tanperature increase in the active
zaie created the conditions for s~zirconium and other exthermic
chemical reactions. Witnesses observed their appearance in the fonn of
fireworks of flying hot and glc:Ming fragrrents.
A mixture of gases containing hydrogen and carbon rronmqde capable of
thexrnal explosion in mixing with air oxygen was foore:i as a result of these
reactions. This mixing could occur after unsealing of ·the reactor space.

As the analysis presented. arove demonstrated, the accident at the
fourth unit of the ChAES belongs to the class of accidents involved with
introduction of excess reactance. The design of the reaction installation
included protection agamst accidents of this type with consideration for
the physical fea’blres of the reactor, including the positive steam
coefficient of reactance.
‘!he technical protection facilities include systems for control and
protection of the reactor against a pc:Mer excess and a decrease in the
runaway period, blocking and protection against malfunctions or switching
.. of the equiprent and systems of the power unit, and a system for errergency
b:X>ling of the reactor •.
Strict rules and an order for conducting the operating process at the
AES, defined by power unit operating regulations, were also included in
addition to the technical protection facilities. Requirements concerning
the unacceptability of a decrease in the operating reactance resezve below
30 rods are am::>ng the m::>st rules.
In the process of preparing for and conducting tests of a
turbogenerator in a rundo.m m:xle with a load of systsn auxiliaries of the
unit, the personnel disengaged a number of technical protection devices and
violated the inp:>rtant conditions of the operating regulations in the
section of safe perfonnance of the operating process.
‘!be table presents a list of the IIDst dangerous violations of
operating conditions ccmnitted by personnel of the fourth unit of the

No. Violation .• M:>tivation Results
1 Decrease in the Attempt to get rut Elrergency·protection
operareactance of “iodine pit” of reactor proved
reserve ineffective
significantly belc:M
the acceptable value
2 Power dip belc:M Operator error in Reactor proved to be
value envisaged by disengagement of I.AR in hard-to-control
testing program state
3 Connection of all Fulf illnent of Temperature of heat-
GTsN to reactor with requirements of transfer medimn of
exceeding of flow testing program KMP’l’s cane close to
rates established by saturation
regulations in t.arp3.rature
regard to individual
4 Blocking of reactor Intention to repeat Loss of possibility
. protection on signal experiment with of autanatic
for shutdcMn of ~ disengagem3Ilt of ‘1’G shutdown of reactor
TG if necessary .5 Blocking of Attempt to conduct Protection of
protection in regard tests despite reactor in regard to
to water level and unstable operation thennal pararreters
steam pressure in of reactor was disengaged
separator drum
·::6 Disengagement of Att:arpt. to avoid Ioss of possibility
system for false response of of reducing scale of
protection against SADR during accident
maxim.mt theoretical perf onnance of
failure testing
(disengagement of
The basic m:>tive in the behavior of the t:ersonnel was the att.arpt to
ccmplete the tests nore quickly. Violation of the established order in
preparation for and perfoi:mance of the tests, violation of the testing
progr~ itself and carelessness in a:mtrol of the reactor installation
attest to inade::jllate understanding ai the part of the personnel of the
features of accarplishment of operating processes in a nuclear reactor and
to their loss of a sense· of the danger.
‘!be developers of the reactor installation did not envisage the
creation of .protective safety systems capable of preventing an
accident in the presence of the set of praredita.ted diversions of technical
protection facilities and violations of cperating regulations which
occurred, since they considered such a set of events impossible.
An ext.rarely ilti>rabable canbination of procedure violations and.
operating conditions tolerated by personnel of the power unit thus was the
original cause of ti:1e accident.
‘!he accident took on catastrophic cllinensions in connection with the
.fact that the reactor was brought by the personnel to a condition so
.contrary to regulations that the effect of a positive reactance coefficient
on the power wild-up was intensified significantly. •

A decisic:n has been made to reset terminal breakers of oontrol rods on.
‘WOrldng nuclear power plants with RBMK reactors such that in the outeJ:m:>st
positioo all rods are inserted into the core to a depth of 1.2 m. This
measure increases the response efficiency of protection and precludes the
possibility of the nultiplication properties of the core fran increasing in·
its lower part when the rod noves fran the upper end piece •. At the sane
tirce a number of absorber rods constantly in the core increases to 70 – 80;
this reduces the steam void effect of reactivity to an allowable value.
· This is a tarqx:>raxy rrea..sure and in the future it will be replaced by
converting RBMK reactors to fuel with initial enrichrren:t 2.4% and placing
additional abf?Qrbers in the core which ensure that positive·coastdCMn of
reactivity not exceed nore than one beta for any change in coolant density.
A number of additional signallers of the cavitation reserve of reactor
coolant p.mps and an autanatic system for canputing reactivi~’ reserve with
ou~t of an E!!rergency reactor shutdown signal when the reserve drops belCM
a given level are being installed. These rreasures have a sarewhat adverse
effect on econanic indicators of nuclear power plants with RBMK, rut
guarantee the necessary safety.
In addition to technical ireasures organizational ones to strengthen
plant discipline and increase operating quality are being int>lenented.
6.1 Fire Fighting on a Nuclear Pa..ter Plant
‘!he primary task after a reactor accident was to control the fire.
As a result of explosions in the reactor an ejection of core fragments
heated to high temperature onto the rooves of certain blildings of reactor
section sexvices, the deaerator, stack and t:mbine roan Il’Ore than 30 f~es
were started. Due to damage to individual oil lines, short circuits in
electrical cables and intense theJ:Ina.l radiation fran the reactor fire foci
were fo~ in the turbine roan above ‘ffi No. 7 Jin the reacto~ roan and the
partially destroyed carparl:m3nts adjacent to it.
At one hcur 30 minutes, fire fighting units for rruclear power plant
protection fran the cities of Pripyat ,. and Chernobyl arrived.
Due to the direct threat of the fire spreading over the cover of the
turbine roc.m to the adjacent third unit and its rapid intensification,
primary rreasures were directed at eliminating the fire in this sector .•
Fires arising within ccmpartrrents were fought using fire extinguishers and
inside stationary fire cranes. By 2 hours 10 minutes mst of the fires had
been put a.it on the roof of the turbine roan and by 2 hc:xlrs 30 minutes on
the roof of the reactor bri.lding. By 0500 the fire had been p.it a.it.
6.2 Estimating fuel condition after the accident
——~– —–·

‘nle accident led to partial.destruction of the reactor core.and
ccnplete destruction of its cooling ·system. Under these condi tians, the ·
state of the environnent in the reactor shaft was detel:mined by the
following processes:
– residual heat release of the fuel due to decay of fission products
– heat release due to different che!llical.reactions taking place in the
reactor shaft (hydrogen canblstion, graphite and zirconium oxidation, etc.);
– heat discharge fran the reactor shaft.due to its cxx:>ling by flows of
at:Irospheric air through holes formed in sealed (before the accident) shells
surrounding the core •
To solve the problem of preventing accident developtent and Umiting
. _its consequences, ·during the first hours after the accident major efforts
were devoted to estimating the fuel state and its possil>le change as time
passed. · To do this, the following analyses had to be done:
– estimate p::>ssible scales of melting (due to residual heat release) of
fuel in the reactor shaft;
– study processes of the interaction of rolten. fuel with reactor
structural materials and reactor shaft materials (mertals, ooncrete and so
– estimate the possibility of melting of construction materials of the
reactor and the shaft due to heat release fran the fuel.·
Initially canputations were done to estimate fuel state in the reactor
shaft with allowance for leakage “Of fission prcxiucts (PD) depending on time
since the accident began.
Study of the dynamics of PD discharge fran the reactor during the first
few days after the accident showed that the fuel tarrperature change as time
passed was naruronotonic. It can be assumed that there were several stages
in the tanperature node of the fuel. The fuel heated up at the instant of
explosion. Tatperature estimation frar1 the anount of relative leakage
(fraction of the isotope.discharging fran the fuel fran its total content in
the fuel at a given point in tine) of iodine radionuclides showed that “the
effective terrperature of the fuel ranaining in the reactor building after
the explosion was 1600 – 1800 K. During the next several dozen minutes,
fuel taaperature droi;:::ped due to release of heat to the graphite·structure
aiid reactor structures. This led to a drop in leakage of volatile PD fran
the fuel.
Here it was considered that the anount of PD discharge fran the reactor
shaft was detennined during this time mainly by processes of graphite
canl::ustion and associated processes of migration of finely dispersed fuel
and PD introduced into the graphite by the accident explosion in the
reactor. Subsequently, the tanperature of the fuel due to residual heat
release began to rise. As a result, leakage of volatile radionuclides
(inert gases, iodine, tellurium, cesium) fran the fuel increased. With the
subsequent tanperature increase of the fuel .leakage of other so-called
nonvolatile radianuclides began~ By· 4 – 5 May, the effective temperature of
the fuel rana.i.ni.ng in the reactor unit stabilized and then began to drop·.
‘!he results of theoretical analyses of fuel state are shown in Fig. 5
‘Which lists results which characterize residual radionuclide content in the
fuel and also the ~ature change of the fuel with allc:Mance for leakage
of PD fran it depending on the tine since the accident began.
Canputations showed:
– maximum fuel temperature cannot reach its m:lting p::>int;
– the PD arerges onto the fuel circuits in batches; .th.is can lead only
to local heatup on the fuel-environrrent boundazy.
The PD escaping fran the fuel fall on structural and other materials
surrounding the reactor in the reactor unit according to condensation and
precipitation temperatures of the fuel. Here radionuclides of Ja:ypton and
zenon escape fran the reactorunit alnost carpletely, the volatile PD
(iodine, cesium) to sare extent and the others remain alnost entirely within
the reactor wilding.
‘!bus the energy of the PD is dissipated through:::>ut the velum: of the
reactor unit.
As the result of these factors nelt:il’lg of the medium surrounding the
fuel and fuel novem:nt becane of ·low probability.
6G3. Limit:ing the Accident Consequences in the Reactor Core
‘!he potential of concentrating part of the nolten fuel·and establishing
conditions for formation of critical mass and a self-containing cha.in
reaction requi.Ied measures against this danger. In addition, the destroyed
reactor was a source of emissions of a large am:::>unt of radioactivity into
the environment.
Irrmediately after the accident, an at~t was made to reduce the
temperature in the reactor shaft and prevent canbustion of the graphite
structure using energency and auxiliacy f eedwater ?Jnq?S to s\lpply water to
the core space. This attempt was unsuccessful.
Innediately one of tw:> decisions had to be made:
– IDcalize the focus of the accident by filling the reactor shaft with
heat discharging and filtering materialsi
– Allow canbustion processes :in the reactor shaft to end naturally.
‘lbe first option was taken since in the second the danger of
radioactive damage to considerable areas with the threat to the health of
the populations of large cities arose. •
A group of specialists in military helicopter