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The accident at Chernobyl Unit 4, on 26 April 1986, did not occur during normal operation of the reactor. It happened during a test designed to assess the reactor's safety margin in a particular set of circumstances. The test, which had to be performed at less than full reactor power, was scheduled to coincide with a routine shut-down of the reactor.


The Test

Nuclear power stations not only produce electricity, they also consume electricity, for example to power the pumps that circulate the coolant. This electricity is usually supplied from the grid. If the source of electricity should fail, most reactors are able to derive the required electricity from their own production. However, if the reactor is operating but not producing power, for example when in the process of shutting down, some other source of supply is required. Generators are generally used to supply the required power, but there is a time delay while they are started.

The test carried out at Chernobyl-4 was designed to demonstrate that a coasting turbine would provide sufficient power to pump coolant through the reactor core while waiting for electricity from the diesel generators. The circulation of coolant was expected to be sufficient to give the reactor an adequate safety margin.

Main Factors in the Accident

Simplified sequence of events

A number of reports have been published that have given summaries of the events leading up to the accident. Since the reactor was destroyed, these summaries have been based on interpretation of evidence. They have not been consistent. There are three reasons for this:

  1. Different researchers have interpreted the same evidence in different ways.
  2. With the passage of time more evidence has become available.
  3. Some authors of reports have been biased.

The sequence of events which follows has been compiled following a review of a large number of reports and it represents what we consider to be the most likely sequence of events.


April 25: Prelude


The scheduled shutdown of the reactor started. Gradual lowering of the power level began .


Lowering of reactor power halted at 1600 MW(t).


The emergency core cooling system (ECCS) was isolated (part of the test procedure) to prevent it from interrupting the test later.
The fact that the ECCS was isolated did not contribute to the accident; however, had it been available it might have reduced the impact slightly.
The power was due to be lowered further; however, the controller of the electricity grid in Kiev requested the reactor operator to keep supplying electricity to enable demand to be met. Consequently, the reactor power level was maintained at 1600 MW(t) and the experiment was delayed.
Without this delay, the test would have been conducted during `day shift'.


Power reduction recommenced.


Shift change.

April 26: Preparation for the test


Power level had been decreased to 720 MW(t) and continued to be reduced.
It is now recognized that the safe operating level for a pre-accident configuration RBMK was about 700 Mwt because of the positive void coefficient.


Power level was now 500 MW(t).
Control was transferred from the local to the automatic regulating system. Either the operator failed to give the `hold power at required level' signal or the regulating system failed to respond to this signal. This led to an unexpected fall in power, which rapidly dropped to 30 MW(t).


In response, the operator retracted a number of control rods in an attempt to restore the power level.
Station safety procedures required that approval of the chief engineer be obtained to operate the reactor with fewer than the effective equivalent of 26 control rods. It is estimated that there were less than this number remaining in the reactor at this time.


The reactor power had risen to 200 MW(t).


An additional pump was switched into the left hand cooling circuit in order to increase the water flow to the core (part of the test procedure).


An additional pump was switched into the right hand cooling circuit (part of the test procedure).
Operation of additional pumps removed heat from the core more quickly. This reduced the water level in the steam separator.


Automatic trip systems to the steam separator were deactivated by the operator to permit continued operation of the reactor.


Operator increased feed water flow in an attempt to address the problems in the cooling system.


Some manual control rods withdrawn to increase power and raise the temperature and pressure in the steam separator.
Operating policy required that a minimum effective equivalent of 15 manual control rods be inserted in the reactor at all times. At this point it is likely that the number of manual rods was reduced to less than this (probably eight). However, automatic control rods were in place, thereby increasing the total number.


Feed water flow rate reduced to below normal by the operator to stabilise steam separator water level, decreasing heat removal from the core.


Spontaneous generation of steam in the core began.


Indications received by the operator, although abnormal, gave the appearance that the reactor was stable.

The Test


Turbine feed valves closed to start turbine coasting. This was the beginning of the actual test.


Automatic control rods withdrawn from the core. An approximately 10 second withdrawal was the normal response to compensate for a decrease in the reactivity following the closing of the turbine feed valves.
Usually this decrease is caused by an increase in pressure in the cooling system and a consequent decrease in the quantity of steam in the core. The expected decrease in steam quantity did not occur due to reduced feed water to the core.


Steam generation increased to a point where, owing to the reactor's positive void coefficient, a further increase of steam generation would lead to a rapid increase in power.


Steam in the core begins to increase uncontrollably.


The emergency button (AZ-5) was pressed by the operator. Control rods started to enter the core.
The insertion of the rods from the top concentrated all of the reactivity in the bottom of the core.


Reactor power rose to a peak of about 100 times the design value.


Fuel pellets started to shatter, reacting with the cooling water to produce a pulse of high pressure in the fuel channels.


Fuel channels ruptured.


Two explosions occurred. One was a steam explosion; the other resulted from the expansion of fuel vapor.

The explosions lifted the pile cap, allowing the entry of air. The air reacted with the graphite moderator blocks to form carbon monoxide. This flammable gas ignited and a reactor fire resulted.

Some 8 of the 140 tones of fuel, which contained plutonium and other highly radioactive materials (fission products), were ejected from the reactor along with a portion of the graphite moderator, which was also radioactive. These materials were scattered around the site. In addition, caesium and iodine vapors were released both by the explosion and during the subsequent fire.

Megawatts - electrical and thermal
The energy produced by nuclear reactors and thermal power stations, is in the form of heat, measured as megawatts thermal - MW(t). The heat is then used to create steam which in turn is used to produce electricity from a generator connected to a steam turbine. The electrical output is measured as MWe (megawatts electric) which is usually about one third of the thermal power rating, depending on the efficiency of the power plant involved.

Chernobyl - Main factors in the accident

  1. Non-routine operation of the reactor.
  2. Violation of operating regulations, including the removal of most of the control rods.
  3. Positive void coefficient characteristic of the reactor.
  4. Apparent lack of knowledge by the station staff of the characteristics of the reactor.
  5. Inadequate control rod design.

Chernobyl - Positive Void Coefficient
Positive void coefficient is a term often associated with the RBMK reactors, the type involved in the Chernobyl disaster. Reactors that have a positive void coefficient can be unstable at low power and may experience a rapid, uncontrollable power increase. While reactors other than the RBMK type have positive void coefficients, they incorporate design features to prevent instability from occuring.

In a water cooled reactor steam may accumulate to form pockets, known as voids. If excess steam is produced, creating more voids than normal, the operation of the reactor is disturbed, because

  1. the water is a more efficient coolant than steam
  2. the water acts as a moderator and neutron absorber while steam does not.

A reactor is said to have a positive void coefficent if excess steam voids lead to increased power generation, and a negative void coefficient if excess steam voids leads to a decrease in power. The coefficient is simply a measure of the speed of change of state of the reactor.

When the void coefficient is positive, the power can increase very rapidly because any power increase that occurs leads to increased steam generation, which in turn leads to a further increase in power. Such increases are, therefore, very difficult to control.

When the void coefficient is negative, excess steam generation will tend to shut down the reactor. This is, of course, not a safety problem.

Most of the world’s operating power reactors have negative void coefficients. In those reactors where same water circuit acts as both moderator and coolant, excess steam generation reduces the slowing of neutrons necessary to sustain the nuclear chain reaction. This leads to a reduction in power.

In some reactor designs however, the moderator and coolant are in separate circuits, or are of different materials. In these reactors, excess steam reduces the cooling of the reactor, but as the moderator remains intact the nuclear chain reaction continues.

In some of these reactors, most notably the RBMK, the neutron absorbing properties of the cooling water are a significant factor in the operating characteristics. In such cases, the reduction in neutron absorbtion as a result of steam production, and the consequent presence of extra free neutrons, enhances the chain reaction. This leads to excess power production.

This excess power production causes additional heating. The additional heat raises the temperature in the cooling circuit and more steam is produced. More steam means less cooling and less neutron absorbtion, and the problem gets worse.

All of this can happen very rapidly. If it is not stopped, and it is very difficult to stop because it feeds itself, there will be the sort of event that happened at Chernobyl unit 4.

In order to avoid problems with positive void coefficient there are two approaches. Either the reactor characteristics can be altered to reduce the positive void coefficient or systems can be provided that will shut the reactor down very quickly if an increase in power is detected.

Since the Chernobyl disaster, the RBMK reactor design has been altered and units have been backfitted to protect them against the effects of the positive void coefficient.

Chernobyl - Post Accident Changes to the RBMK

Immediate Safety Changes
After the accident at Chernobyl unit 4, the primary concern was to reduce the positive void coefficient. All operating RBMK reactors, in the former Soviet Union therefore, had the following changes implemented to improve operating safety:

  1. To improve the operational reactivity margin the effective number of manual control rods was increased from 30 to 45.
  2. The installation of 80 additional absorbers in the core to inhibit operation at low power.
  3. An increase in fuel enrichment from 2% to 2.4% to maintain fuel burnup with the increase in neutron absorption.

These factors have reduced the positive void coefficient from +4.5Beta [Greek symbol] to +0.7Beta [Greek symbol], eliminating the possibility of power excursion. Beta [Greek symbol] is the delayed neutron fraction, which is neutrons emitted with a measurable time delay. The next consideration was to reduce the time taken to shut the reactor down and eliminate the positive void reactivity. Improvements include:

  1. Scram (shut down) rod insertion time cut from 18 to 12 seconds.
  2. The redesign of control rods.
  3. The installation of a fast scram system (-2Beta [Greek symbol]/2.5s).
  4. Precautions against unauthorised access to emergency safety systems.

RBMK Modifications
In addition to the safety changes, it has been recommended that RBMKs are modified, a procedure which is currently on going at the Leningrad site. Chernobyl unit 1 was licensed for operation in October 1995, following extensive maintainance which included the removal of some fuel channels to evaluate the metal and some backfitting. The modification process is commonly referred to as backfitting and consists of:

  1. Replacement of the fuel channels at all units except Smolensk-3.
  2. Replacement of the group distribution headers and addition of check valves.
  3. Improvements to the emergency core cooling systems.
  4. Improvements of the reactor cavity over-pressure protection systems.
  5. Replacement of the process computer, SKALA.
RBMK - Control Rod Redesign

One of the post-accident changes to the RBMK was the redesign of the control rods.

179 of 211 control rods are inserted into the core from the top. To improve their effectiveness, they are equipped with "riders" fixed to their bottom end but with a gap between the rider and the bottom tip of the control rod. Approximately 1.0m water columns remained under and above it. When the control rod is in its uppermost position, the rider is in the control rod cooling tube within the fuelled region of the core. The rider being made substantially of graphite, is almost transparent to neutrons, while water, which would occupy the tube otherwise, plays as an absorber. When the reactor is "poisoned" with Xenon and with partially inserted control rods, the major part of the power is produced within the lower region of the core. This means that when the rod started to move down from its uppermost position, the rider removed water from the lower part, causing an increase in reactivity and hence in power.
It was calculated that a surge of reactivity after the emergency shut down button had been pressed could reach +2Beta[Greek symbol]eff. To counter this problem the control rod design has been changed with the rider tie length increased to prevent water columns in the lower part of the core.

RBMK light-water graphite reactor

The Soviet designed RBMK is a pressurised water reactor with individual fuel channels and using ordinary water as its coolant and graphite as its moderator. It is very different from most other power reactor designs as it was intended and used for both plutonium and power production. The combination of graphite moderator and water coolant is found in no other power reactors. The design characteristics of the reactorwere shown, in the Chernobyl accident, to cause instability when low power. This was due primarily to control rod design and a positive void coefficient. A number of significant design changes have now been made to address these problems.
RBMK Reactor
Fuel rods Pellets of enriched uranium oxide are enclosed in a zircaloy tube 3.65m long, forming a fuel rod. Two sets of 18 fuel rods are arranged cylindrically in a carriage to form a fuel assembly of about 10 m length. These fuel assemblies can be lifted into and out of the reactor mechanically, allowing fuel replenishment while the reactor is in operation.
Pressure tubes Within the reactor each fuel assembly is positioned in its own pressure tube or channel. Each channel is individually cooled by pressurised water.
Graphite moderator A series of graphite blocks surround, and hence separate, the pressure tubes. They act as a moderator to slow down the neutrons released during fission. This is necessary for continuous fission to be maintained. Conductance of heat between the blocks is enhanced by a mixture of helium and nitrogen gas.
Control rods Boron carbide control rods absorb neutrons to control the rate of fission. A few short rods, inserted upwards from the bottom of the core, even the distribution of power across the reactor. The main control rods are inserted from the top down and provide automatic, manual or emergency control. The automatic rods are regulated by feedback from in-core detectors. If there is a deviation from normal operating parameters (e.g. increased reactor power level), the rods can be dropped into the core to reduce or stop reactor activity. A number of rods normally remain in the core during operation.
Coolant Two separate water coolant systems each with four pumps circulate water through the pressure tubes. Ninety-five percent of the heat from fission is transferred to the coolant. There is also an emergency core cooling system which will come into operation if either coolant circuit is interrupted.
Steam Separator Steam from the heated coolant is fed to turbines to produce electricity in the generator. The steam is then condensed and fed back into the circulating coolant.
Containment The reactor core is located in a concrete lined cavity that acts as a radiation shield. The upper shield or pile cap above the core, is made of steel and supports the fuel assemblies. The steam separators of the coolant systems are housed in their own concrete shields.
Reactor Core Enlargement