This article is about the investigations into and trials related to the Chernobyl nuclear disaster. For more general information, see Chernobyl disaster.
Investigations into the Chernobyl nuclear accident
The IAEA produced two significant reports on Chernobyl; INSAG-1 in 1986, and a revised report, INSAG-7 in 1992. In summary, according to INSAG-1, the main cause of the accident was the operators' actions, but according to INSAG-7, the main cause was the reactor's design.[2]:24[3] Both IAEA reports identified an inadequate "safety culture" (INSAG-1 coined the term) at all managerial and operational levels as a major underlying factor of different aspects of the accident. This was stated to be inherent not only in operations but also during design, engineering, construction, manufacture and regulation.[2]:21,24
Views of the main causes were heavily lobbied by different groups, including the reactor's designers, power plant personnel, and the Soviet and Ukrainian governments. This was due to the uncertainty about the actual sequence of events and plant parameters. After INSAG-1 more information became available, and more powerful computing has allowed better forensic simulations.[2]:10
The INSAG-7 conclusion of major factors contributory to the accident was:
"The Accident is now seen to have been the result of concurrence of the following major factors: specific physical characteristics of the reactor; specific design features of the reactor control elements; and the fact that the reactor was brought to a state not specified by procedures or investigated by an independent safety body. Most importantly, the physical characteristics of the reactor made possible its unstable behaviour."[2]:23
INSAG-1 report (1986)
The first official Soviet explanation of the accident was given by Soviet scientists and engineers to representatives of IAEA member states and other international organisations at the first Post-Accident Review Meeting, held at the IAEA in Vienna 25–29 August 1986. This explanation effectively placed the blame on the power plant operators. The IAEA INSAG-1 report followed shortly afterwards in September 1986, and on the whole also supported this view, based also on the information provided in discussions with the Soviet experts at the Vienna review meeting.[4] In this view, the catastrophic accident was caused by gross violations of operating rules and regulations. For instance; "During preparation and testing of the turbine generator under run-down conditions using the auxiliary load, personnel disconnected a series of technical protection systems and breached the most important operational safety provisions for conducting a technical exercise."[5]:311
It was stated that at the time of the accident the reactor was being operated with many key safety systems turned off, most notably the emergency core cooling system (ECCS), LAR (Local Automatic control system), and AZ (emergency power reduction system). Personnel had an insufficient understanding of technical procedures involved with the nuclear reactor, and knowingly ignored regulations to expedite the electrical test completion.[5] Several procedural irregularities also helped to make the accident possible, one of which was insufficient communication between the safety officers and the operators in charge of the test.[citation needed]
It was held that the designers of the reactor considered this combination of events to be impossible and therefore did not allow for the creation of emergency protection systems capable of preventing the combination of events that led to the crisis, namely the intentional disabling of emergency protection equipment plus the violation of operating procedures. Thus the primary cause of the accident was the extremely improbable combination of rule infringement plus the operational routine allowed by the power station staff.[5]:312
On the disconnection of safety systems, Valery Legasov said in 1987, "It was like airplane pilots experimenting with the engines in flight."[6] In this analysis the operators were blamed, but deficiencies in the reactor design and in the operating regulations that made the accident possible were set aside and mentioned only casually. This view was reflected in numerous publications and artistic works on the theme of the Chernobyl accident that appeared immediately after the accident,[7] and for a long time remained dominant in the public consciousness and in popular publications.
The trial took place from 7 to 30 July 1987 in a temporary courtroom set up in the House of Culture in the city of Chernobyl, Ukraine. Five plant employees (Anatoly S. Dyatlov, the former deputy chief engineer; Viktor P. Bryukhanov, the former plant director; Nikolai M. Fomin, the former chief engineer; Boris V. Rogozhin, the shift director of Reactor 4; and Aleksandr P. Kovalenko, the chief of Reactor 4); and Yuri A. Laushkin (Gosatomenergonadzor [USSR State Committee on Supervision of Safe Conduct of Work in Atomic Energy] inspector) were sentenced to ten, ten, ten, five, three, and two years respectively in labor camps.[8] The families of Aleksandr Akimov, Leonid Toptunov and Valery Perevozchenko had received official letters, but prosecution against the employees had been terminated at their deaths.
Anatoly Dyatlov was found guilty "of criminal mismanagement of potentially explosive enterprises" and sentenced to ten years imprisonment—of which he would serve three[9]—for the role that his oversight of the experiment played in the ensuing accident.
INSAG-7 report (1992)
Reactor hall No. 1 of the Chernobyl PlantA simplified diagram comparing the Chernobyl RBMK and the most common nuclear reactor design, the Light water reactor. RBMK issues: 1. Using a graphite moderator in a water-cooled reactor, permitting criticality in a total loss of coolant accident. 2. A positive steam void coefficient that made the destructive power excursion possible. 3. Control rods design; taking 18–20 seconds to be fully inserted, and with graphite tips that increased reactivity initially. 4. No reinforced containment building.
In 1991 a Commission of the USSR State Committee for the Supervision of Safety in Industry and Nuclear Power reassessed the causes and circumstances of the Chernobyl accident and came to new insights and conclusions. Based on that, INSAG published an additional report, INSAG-7,[2] which reviewed "that part of the INSAG-1 report in which primary attention is given to the reasons for the accident," and this included the text of the 1991 USSR State Commission report translated into English by the IAEA as Annex I.[2]
By the time of this report, the post-Soviet Ukrainian government had declassified a number of KGB documents from the period between 1971 and 1988 related to the Chernobyl plant. It mentioned, for example, previous reports of structural damage caused by negligence during construction of the plant (such as splitting of concrete layers) that were never acted upon. They documented more than 29 emergency situations in the plant during this period, eight of which were caused by negligence or poor competence on the part of personnel.[12]
In the INSAG-7 report, most of the earlier accusations against staff for breach of regulations were acknowledged to be either erroneous, being based on incorrect information obtained in August 1986, or were judged less relevant. The INSAG-7 report also reflected the view of the 1991 USSR State Commission account which held that the operators' actions in turning off the emergency core cooling system, interfering with the settings on the protection equipment, and blocking the level and pressure in the separator drum did not contribute to the original cause of the accident and its magnitude, although they may have been a breach of regulations. In fact, turning off the emergency system designed to prevent the two turbine generators from stopping was not a violation of regulations.[2] Soviet authorities had identified a multitude of operator actions as regulation violations in the original 1986 report while no such regulations were in fact in place.[2]:18
The primary design cause of the accident, as determined by INSAG-7, was a major deficiency in safety features,[2]:22 in particular the "positive scram" effect due to the control rods' graphite tips that actually initially increased reactivity when control rods entered the core to reduce reactivity.[2]:16 There was also an overly positive void coefficient of the reactor, whereby steam-generated voids in the fuel cooling channels would increase reactivity because neutron absorption was reduced, resulting in more steam generation, and thereby more voids; a regenerative process.[2]:13
To avoid such conditions, it was necessary for the operators to track the value of the reactor operational reactivity margin (ORM) but this value was not readily available to the operators[2]:17 and they were not aware of the safety significance of ORM on void and power coefficients.[2]:14 However, regulations did forbid operating the reactor with a small margin of reactivity. Yet "post-accident studies have shown that the way in which the real role of the ORM is reflected in the Operating Procedures and design documentation for the RBMK-1000 is extremely contradictory", and furthermore, "ORM was not treated as an operational safety limit, violation of which could lead to an accident".[2]:34–25
Even in this revised analysis, the human factor remained identified as a major factor in causing the accident; particularly the operating crew's deviation from the test programme. "Most reprehensibly, unapproved changes in the test procedure were deliberately made on the spot, although the plant was known to be in a very different condition from that intended for the test."[2]:24 This included operating the reactor at a lower power level than the prescribed 700MW before starting the electrical test. The 1986 assertions of Soviet experts notwithstanding, regulations did not prohibit operating the reactor at this low power level.[2]:18
INSAG-7 also said, "The poor quality of operating procedures and instructions, and their conflicting character, put a heavy burden on the operating crew, including the chief engineer. The accident can be said to have flowed from a deficient safety culture, not only at the Chernobyl plant, but throughout the Soviet design, operating and regulatory organizations for nuclear power that existed at that time."[2]:24
Positive void coefficient
The reactor had a dangerously large positive void coefficient of reactivity. The void coefficient is a measurement of how a reactor responds to increased steam formation in the water coolant. Most other reactor designs have a negative coefficient, i.e. the nuclear reaction rate slows when steam bubbles form in the coolant, since as the steam voids increase, fewer neutrons are slowed down. Faster neutrons are less likely to split uranium atoms, so the reactor produces less power (negative feedback effect).[2]
Chernobyl's RBMK reactor, however, used solid graphite as a neutron moderator to slow down the neutrons, and the cooling water acted as a neutron absorber. Thus, neutrons are moderated by the graphite even if steam bubbles form in the water. Furthermore, because steam absorbs neutrons much less readily than water, increasing the voids means that more moderated neutrons are able to split uranium atoms, increasing the reactor's power output. This could create a positive feedback regenerative process (known as a positive power coefficient) which makes the RBMK design very unstable at low power levels, and prone to sudden energy surges to a dangerous level. Not only was this behaviour counter-intuitive, this property of the reactor under certain conditions was unknown to the personnel.[2]
Control rod design
There was a significant flaw in the design of the control rods. The reactor core was 7 metres (23 feet) high. The upper half of the rod 7 metres (23 feet) was boron carbide, which absorbs neutrons and thereby slows the reaction. The bottom section of each control rod was a 4.5 meter graphite displacer, which prevented the channels from filling with water when rods were withdrawn. The flaw lay in the 1.25-metre (4.1-foot) gap between the bottom of the graphite displacer and the bottom of the reactor, meaning that the lowest portion of control rod channel was filled with water and not graphite. See page 123. Fig 11–10.[2] With this design, when the rods were inserted from the fully retracted position to stop the reaction on the AZ-5 signal, the graphite displaced neutron-absorbing water, causing fewer neutrons to be absorbed and increasing reactivity. For the first 11 to 14 seconds of rod deployment until the boron was in position, reactor power across the floor of the reactor could increase, rather than decrease. This feature of control rod operation was counter-intuitive and not known to the reactor operators.
Management and operational deficiencies
Other deficiencies were noted in the RBMK-1000 reactor design, as were its non-compliance with accepted standards and with the requirements of nuclear reactor safety. While INSAG-1 and INSAG-7 reports both identified operator error as an issue of concern, the INSAG-7 identified that there were numerous other issues that were contributing factors that led to the incident. These contributing factors include:
The plant was not designed to safety standards in effect and incorporated unsafe features
A nuclear reactor is a device used to initiate and control a fission nuclear chain reaction or nuclear fusion reactions. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid, which in turn runs through steam turbines. These either drive a ship's propellers or turn electrical generators' shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating. Some reactors are used to produce isotopes for medical and industrial use, or for production of weapons-grade plutonium. As of 2022, the International Atomic Energy Agency reports there are 422 nuclear power reactors and 223 nuclear research reactors in operation around the world.
A pressurized water reactor (PWR) is a type of light-water nuclear reactor. PWRs constitute the large majority of the world's nuclear power plants. In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated, high pressure water then flows to a steam generator, where it transfers its thermal energy to lower pressure water of a secondary system where steam is generated. The steam then drives turbines, which spin an electric generator. In contrast to a boiling water reactor (BWR), pressure in the primary coolant loop prevents the water from boiling within the reactor. All light-water reactors use ordinary water as both coolant and neutron moderator. Most use anywhere from two to four vertically mounted steam generators; VVER reactors use horizontal steam generators.
A nuclear meltdown is a severe nuclear reactor accident that results in core damage from overheating. The term nuclear meltdown is not officially defined by the International Atomic Energy Agency or by the United States Nuclear Regulatory Commission. It has been defined to mean the accidental melting of the core of a nuclear reactor, however, and is in common usage a reference to the core's either complete or partial collapse.
The RBMK is a class of graphite-moderated nuclear power reactor designed and built by the Soviet Union. The name refers to its design where, instead of a large steel pressure vessel surrounding the entire core, the core is surrounded by a cylindrical annular steel tank inside a concrete vault and each fuel assembly is enclosed in an individual 8 cm (inner) diameter pipe. The channels also contain the coolant, and are surrounded by graphite.
A scram or SCRAM is an emergency shutdown of a nuclear reactor effected by immediately terminating the fission reaction. It is also the name that is given to the manually operated kill switch that initiates the shutdown. In commercial reactor operations, this type of shutdown is often referred to as a "scram" at boiling water reactors (BWR), a "reactor trip" at pressurized water reactors and EPIS at a CANDU reactor. In many cases, a scram is part of the routine shutdown procedure, which serves to test the emergency shutdown system.
A loss-of-coolant accident (LOCA) is a mode of failure for a nuclear reactor; if not managed effectively, the results of a LOCA could result in reactor core damage. Each nuclear plant's emergency core cooling system (ECCS) exists specifically to deal with a LOCA.
In nuclear engineering, the void coefficient is a number that can be used to estimate how much the reactivity of a nuclear reactor changes as voids form in the reactor moderator or coolant. Net reactivity in a reactor is the sum total of multiple contributions, of which the void coefficient is but one. Reactors in which either the moderator or the coolant is a liquid typically will have a void coefficient value that is either negative or positive. Reactors in which neither the moderator nor the coolant is a liquid will have a void coefficient value equal to zero. It is unclear how the definition of "void" coefficient applies to reactors in which the moderator/coolant is neither liquid nor gas.
The light-water reactor (LWR) is a type of thermal-neutron reactor that uses normal water, as opposed to heavy water, as both its coolant and neutron moderator; furthermore a solid form of fissile elements is used as fuel. Thermal-neutron reactors are the most common type of nuclear reactor, and light-water reactors are the most common type of thermal-neutron reactor.
In nuclear engineering, prompt criticality describes a nuclear fission event in which criticality is achieved with prompt neutrons alone and does not rely on delayed neutrons. As a result, prompt supercriticality causes a much more rapid growth in the rate of energy release than other forms of criticality. Nuclear weapons are based on prompt criticality, while nuclear reactors rely on delayed neutrons or external neutrons to achieve criticality.
Passive nuclear safety is a design approach for safety features, implemented in a nuclear reactor, that does not require any active intervention on the part of the operator or electrical/electronic feedback in order to bring the reactor to a safe shutdown state, in the event of a particular type of emergency. Such design features tend to rely on the engineering of components such that their predicted behaviour would slow down, rather than accelerate the deterioration of the reactor state; they typically take advantage of natural forces or phenomena such as gravity, buoyancy, pressure differences, conduction or natural heat convection to accomplish safety functions without requiring an active power source. Many older common reactor designs use passive safety systems to a limited extent, rather, relying on active safety systems such as diesel-powered motors. Some newer reactor designs feature more passive systems; the motivation being that they are highly reliable and reduce the cost associated with the installation and maintenance of systems that would otherwise require multiple trains of equipment and redundant safety class power supplies in order to achieve the same level of reliability. However, weak driving forces that power many passive safety features can pose significant challenges to effectiveness of a passive system, particularly in the short term following an accident.
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The supercritical water reactor (SCWR) is a concept Generation IV reactor, designed as a light water reactor (LWR) that operates at supercritical pressure. The term critical in this context refers to the critical point of water, and should not be confused with the concept of criticality of the nuclear reactor.
The water-water energetic reactor (WWER), or VVER is a series of pressurized water reactor designs originally developed in the Soviet Union, and now Russia, by OKB Gidropress. The idea of such a reactor was proposed at the Kurchatov Institute by Savely Moiseevich Feinberg. VVER were originally developed before the 1970s, and have been continually updated. As a result, the name VVER is associated with a wide variety of reactor designs spanning from generation I reactors to modern generation III+ reactor designs. Power output ranges from 70 to 1300 MWe, with designs of up to 1700 MWe in development. The first prototype VVER-210 was built at the Novovoronezh Nuclear Power Plant.
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A graphite-moderated reactor is a nuclear reactor that uses carbon as a neutron moderator, which allows natural uranium to be used as nuclear fuel.
Anatoly Stepanovich Dyatlov was a Soviet engineer who was the deputy chief engineer for the Chernobyl Nuclear Power Plant. He supervised the safety test which resulted in the 1986 Chernobyl disaster, for which he served time in prison as he was blamed for not following the safety protocols. He was released due to health concerns in 1990.
Leningrad Nuclear Power Plant ) is a nuclear power plant located in the town of Sosnovy Bor in Russia's Leningrad Oblast, on the southern shore of the Gulf of Finland, some 70 kilometres (43 mi) to the west of the city centre of Saint Petersburg.
Shutdown is the state of a nuclear reactor when the fission reaction is slowed significantly or halted completely. Different nuclear reactor designs have different definitions for what "shutdown" means, but it typically means that the reactor is not producing a measurable amount of electricity or heat, and is in a stable condition with very low reactivity.
At its peak in 1982, nuclear power in the Soviet Union accounted for 6.5% of total electricity consumption and the total nuclear capacity installed was 18 GW. However, nuclear power within the Soviet Union declined severely as a result of the 1986 Chernobyl Disaster.
Leonid Fedorovych Toptunov was a Soviet electrical engineer who was the senior reactor control chief engineer at the Chernobyl Nuclear Power Plant Reactor Unit 4 on the night of the Chernobyl disaster, 26 April 1986.
References
↑ "History of the International Atomic Energy Agency", IAEA, Vienna (1997).
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