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TIMESTAMPS

SOVIET NUCLEAR POWER PLANT DESIGNS

Key Facts

• Western-style plants employ the design principle of safety in depth, relying on a series of physical barriers--including a massive reinforced concrete structure called the containment--to prevent the release of radioactive material to the environment. With the exception of the VVER-1000 design, Soviet-designed reactors do not have such a containment structure.
• Soviet-designed reactors are essentially variations on two basic designs: the VVER--or pressurized light water--type, and the RBMK--the graphite moderated, channel reactor.
• Three generations of Soviet-designed VVER reactors--upgraded over time--are operating in Eastern Europe and the former Soviet Union. The first generation--the VVER-440 Model V230--operates at four plant sites in three countries: Russia, Bulgaria and the Slovak Republic. The second generation--the VVER-440 Model V213--operates at five plant sites in five countries: Russia, Ukraine, Hungary, the Czech Republic and the Slovak Republic. The third generation--the VVER 1000--operates at eight plant sites in three countries: Russia, Ukraine and Bulgaria.
• At the time of the collapse of the Soviet Union, two advanced versions of the VVER-1000 were under development. Russia has continued the development of an upgraded VVER-1000, and has developed a new design for a 1000-megawatt reactor with enhanced safety features.
• Three generations of RBMK reactors are operating in the former Soviet Union: 11 units in Russia, two in Ukraine and two in Lithuania. Despite improvements to the RBMK design since the Chernobyl accident, concerns remain about these reactors, especially the first-generation ones.

Major Difference Between Soviet-Designed and Western Nuclear Power Plants

Soviet-designed nuclear power plants differ from Western nuclear power plants in many respects, including plant instrumentation and controls, safety systems and fire protection systems.

While Soviet-designed plants--like Western-style plants--employ the design principle known in the West as "safety in depth," only one reactor design includes a containment structure as part of that principle.

In the unlikely event that safety systems fail, plants designed on the "safety in depth" principle rely on a series of physical barriers to prevent the release of radioactive material to the environment. At U.S. plants:

• The first barrier is the nuclear fuel itself, which is in the form of solid ceramic pellets. Most of the radioactive by-products of the fission process remain bound inside the fuel pellets.
• These pellets are then sealed in rods, made of special steel, about 12 feet long and half an inch in diameter.
• The fuel rods are inside a large steel pressure vessel, which has walls about eight inches thick.
• At most plants, this vessel is enclosed in a large, leak-tight shell of steel plate.
• All this is contained inside a massive concrete structure--called the containment--with walls several feet thick.

Most Soviet-designed reactors employ similar features, but only the VVER-1000 design has a containment structure like that of most nuclear power plants elsewhere in the world. Without this protection, radioactive material could escape to the environment in the event of a serious accident.

Plant Location and Design "Families"

At the end of 1994, more than 60 commercial nuclear reactors of Soviet design were operating or under construction in Russia, Ukraine, Lithuania, Bulgaria, the Czech Republic, the Slovak Republic and Hungary.

A two-unit Soviet-designed nuclear plant in Finland was built using the VVER-440 Model 213 basic design, but was upgraded to include a Western instrumentation and control system and a containment structure.

With the exception of small nuclear units used for district steam heating and several small fast-breeder reactors--which produce fuel as they generate electricity--Soviet-designed commercial nuclear power plants are variations on two basic designs: the VVER--or pressurized light water--type, and the RBMK--the graphite moderated, channel reactor. There are no RBMK plants operating outside the former Soviet Union.

Like all nuclear units based on light water technology, the Soviet VVER design uses water to generate steam and to cool the reactor. Water also acts as a "moderator," slowing down the atomic particles (neutrons) in the nuclear reaction to increase the chances of fissioning, or splitting. The "moderating" effect of the water adds safety, because a water loss will slow the nuclear chain reaction.

In the RBMK design, graphite is used in place of water as a moderator, surrounding vertical pressure tubes which hold the nuclear fuel and the water that will be boiled to steam. Unlike light water units, the RBMK's nuclear chain reaction and power output increase when cooling water is lost. This design flaw--called a "positive void coefficient"--caused the uncontrollable power surge that led to the Chernobyl accident. The corrections and modifications made to all of the RBMKs since the Chernobyl accident are generally considered to be adequate to preclude the type of nuclear excursion--a sudden, rapid rise in power level--that occurred at Chernobyl Unit 4 in April 1986.

The Beloyarsk fast-breeder reactor in Russia is the second-largest such unit in the world, behind the French Super Phenix, and generates new fuel as it operates. The major components of the Beloyarsk unit are submerged in a large pool of liquid sodium, which acts as a moderator and transfers heat away from the reactor to boil water to make steam.

The VVER: Three Generations of Light Water Reactors, Upgraded Over Time

Although it shares a basic engineering concept with its counterparts in the United States, France and Japan, the Soviet pressurized water--or VVER--design is very different and does not meet Western safety standards. However, second and third generation plants of this design--the VVER-440 Model V213 and VVER-1000--are widely viewed as having a design safety basis sufficiently comparable to that used in the West to justify short-term and long-term safety and performance upgrades on both safety and economic grounds. Such upgrades would allow these plants to continue operating at an acceptable level of safety. At the same time, regulatory requirements and the extent of plant upgrading may differ from country to country and plant to plant, resulting in varying levels of safety, even for plants of the same model.

First-Generation VVERs

The earliest pressurized water nuclear plants were developed by the Soviets between 1956 and 1970. These plants include the following versions:

Principal Strengths:

• Six primary coolant loops (providing multiple paths for cooling the reactor), each with a horizontal steam generator (for better heat transfer), which together provide a large volume of coolant. In some respects this design is more forgiving than Western plant designs with two, three or four large vertical steam generators.
• Low power density in the fuel, coupled with the large inventory of primary coolant, provide more time than a Western-style PWR for operator actions before fuel failures would occur.
• Isolation valves that allow plant operators to take one or more of the six coolant loops out of service for repair while continuing to operate the plant. This feature is found in only a few Western plants.
• Ability to sustain a simultaneous loss of coolant and off-site power, due to coolant pumps and two internal power generators that "coast down" after a shutdown.
• Plant worker radiation levels reportedly lower than many Western plants, due to selection of materials, high-capacity primary coolant purification system, and water chemistry control.
• Ability to produce significant amounts of power despite design and operating deficiencies.

Principal Deficiencies:

• Accident Localization System--which serves as a reactor confinement--designed to handle only one four-inch pipe rupture. If larger coolant pipes rupture, this system vents directly to the atmosphere through nine large vent valves. Western nuclear plants have containments designed for rupture of the largest pipes. In addition, the confinement has very small volume, very poor leak-tightness and poor hydrogen mitigation.
• No emergency core cooling systems or auxiliary feedwater systems similar to those required in Western nuclear plants.
• Major concern about embrittlement (gradual weakening) of the reactor pressure vessel surrounding nuclear fuel, due to lack of internal stainless-steel cladding and use of low-alloy steel with high levels of impurities.
• Plant instrumentation and controls, safety systems, fire protection systems, and protection for control room operators below Western standards.
• Quality of materials, construction, operating procedures and personnel training below Western standards.

Second-Generation VVERs

The VVER-440 Model V213 was designed between 1970 and 1980. The development of this design coincided with the first uniform safety requirements drawn up by Soviet designers.

VVER-440 Model V213 units in the former Soviet Union include:

VVER-440 Model V213 units in Central and Eastern Europe include:

Construction of a version of the Model V213 intended for export began in Cuba in 1983 but was suspended in 1992.

Principal Strengths:

• Upgraded Accident Localization System vastly improved over the earlier VVER-440 Model V230 design, comparable to several Western plants, and using a vapor-suppression confinement structure called a "bubbler-condenser" tower.
• Low power density in the fuel, coupled with the large inventory of primary coolant, provides more time than a Western-style PWR for operator actions before fuel failures would occur.
• Addition of emergency core cooling and auxiliary feedwater systems.
• Reactor pressure vessel with stainless-steel internal lining to alleviate much concern about the vessel embrittlement associated with the earlier VVER-440 Model V230 design.
• Improved coolant pump, and continued use of six coolant loops (providing multiple paths for cooling the reactor) and horizontal steam generators (for better heat transfer) with large coolant volume.
• Standardization of plant components, providing extensive operating experience for many parts and making possible incremental improvements and backfits of components.

Principal Deficiencies:

• Plant instrumentation and controls--for example, reactor protection systems and diagnostics--behind Western standards. Significant variations exist among countries with VVER-440 Model V213 plants.
• Separation of plant safety systems (to help assure that an event in one system will not interfere with the operation of others), fire protection, and protection for control room operators improved over Model V230 plants, but generally below Western standards.
• Poor leak-tightness of confinement.
• Unknown quality of plant equipment and construction, due to lack of documentation on design, manufacturing and construction, and reported instances of poor-quality materials being re-worked at plant sites.
• Major variations in operating and emergency procedures, operator training, and operational safety (for example, use of control-room simulators) among plants. These aspects of plant operations depend primarily on the organization or country operating Model V213 plants rather than on the plant supplier. Some countries have added safety features to their Model V213 plants.

Third-Generation VVERs

The VVER-1000 design was developed between 1975 and 1985 based on the requirements of a new Soviet nuclear standard that incorporated some international practices, particularly in the area of plant safety. The VVER-1000 design was intended to be used for many plants, and 18 units now operate in two former Soviet republics:

Ukraine

Balakovo 1-4 Rovno 3 Khmelnitskiy 1 South Ukraine 1-3 Zaporozhye 1-6

Two VVER-1000 units were built outside the former Soviet Union:

Work was stopped on two other VVER-1000 units in Bulgaria (Belene 1 and 2) after public protests over claims of unsuitable soil and seismic conditions.

The Hungarian government canceled Paks 5 and 6 in 1989.

Construction of two VVER-1000 units at Stendal, in the former East Germany, was halted following reunification with West Germany.

Two VVER-1000 units under construction at Temelin in the Czech Republic are being upgraded with Western instrumentation and control equipment and fuel.

A total of 25 VVER-1000 units are at some stage of construction in the former Soviet Union--15 in Russia and 10 in Ukraine. But in 1992, construction on 11 of these units in Russia, and seven in Ukraine, had reportedly been canceled or deferred indefinitely.

Of the VVER-1000 units earmarked for completion under Russia's 20-year nuclear plant construction plan, announced in 1992, Balakovo 4 came on line in 1993, and Kalinin 3--originally scheduled to come on line in 1995--will now reportedly be operational by the end of 1996. The status of two other VVER-1000 units--Balakovo 5 and 6--is uncertain; according to a plant official, the country's new nuclear energy law requires the construction procedure to begin again, with the involvement of local and regional authorities. Rostov 1 is reportedly 90 percent complete and will be finished by the year 2000. Rostov 2 is 40 percent complete, and will be finished soon after Rostov 1.

Principal Strengths:

• Steel-lined, pre-stressed, large-volume concrete containment structure, similar in function to Western nuclear plants.
• "Evolutionary" design incorporating safety improvements over VVER-440 Model V213 plants. The Soviet approach to standardization was based on continued use of components that had performed well in earlier plants.
• Use of four coolant loops and horizontal steam generators--both considered improvements by Soviet designers.
• Redesigned fuel assemblies that allow better flow of coolant, and improved control rods.
• Plant worker radiation levels reportedly lower than in many Western plants, apparently due to selection of materials, high-capacity system for purifying primary coolant, and water chemistry control.

Principal Deficiencies:

• Substandard plant instrumentation and controls. Wiring of emergency electrical system and reactor protection system does not meet Western standards for separation--control and safety functions are interconnected in ways that may allow failure of a control system to prevent operation of a safety system.
• Fire protection systems that do not appear to differ substantially from earlier VVER models, which do not meet Western standards.
• Quality control, design and construction significantly deficient by U.S. standards.
• Protection measures for control-room operators essentially unchanged from earlier VVER-440 Model V213 design, which does not meet U.S. standards. Unlike all U.S. nuclear plants, and most in Western countries, VVER-1000s have no on-site "technical support center" to serve as a command post for stabilizing the plant in an emergency. Technical support centers were incorporated in U.S. and many Western nuclear plants following the accident at Three Mile Island Unit 2 in 1979.
• Operating and emergency procedures that fall far short of Western standards and vary greatly among operators of VVER-1000 plants.

VVER-1000 Derivatives

Even before the breakup of the Soviet Union, derivative versions of the VVER-1000 were under development.

The VVER-88 concept is a basic VVER-1000 with post-Chernobyl improvements. Although it included a number of safety advances, it was not considered economical, and none were built.

In 1987, design work was begun on the VVER-1800, a VVER-1000 upgraded for greater safety and economy. The VVER-1800 design incorporated a lower-power reactor core, annual refueling, and more reliable control and protection systems.

In 1989, Finland and the Soviet Union jointly announced the start of development work on the VVER-91, a VVER-1000 version that would meet stringent Finnish nuclear plant design requirements. On paper, the Soviet VVER-91 design is among the world's most advanced light water nuclear power plants. The People's Republic of China has ordered two 1000-megawatt units--the VVER-91 model--from Russia.

Development of a new VVER-1000 design, the VVER-92, was expected to be carried out with Western assistance. The VVER-92 incorporated what one Finnish nuclear expert called "radically simplified" plant systems that included active safety systems, a reduced-power reactor core, and a double containment structure surrounding the nuclear reactor. However, the Ministry of Atomic Energy has diverted some funding for VVER-92 development to a pilot project for building a smaller advanced VVER, the VVER-640, according to an official of Rosenergoatom, Russia's utility organization.

Russia's 20-year nuclear plant construction plan includes a 1000-megawatt VVER reactor called the NP-1000.

The RBMK: The Chernobyl-Type Soviet Nuclear Power Plant

The former Soviet Union built 17 nuclear units based on the RBMK design used at the Chernobyl nuclear power plant, the site of the world's worst commercial nuclear accident. There are currently 15 RBMK reactors in operation: 11 units in Russia, two in Ukraine and two in Lithuania. These units were connected to the grid between 1973 (Leningrad 1) and 1990 (Smolensk 3). During these 17 years, the design evolved significantly. In addition, following the Chernobyl accident in 1986, some major safety upgrades were implemented. Today it is generally recognized that there are three generations of RBMK nuclear power plants, although even within a given generation the units can differ substantially.

RBMKs in the former Soviet republics include:

RBMKs of the 1,500-megawatt class include:

At the time of the Chernobyl accident, six RBMK units were under construction in the U.S.S.R.: Kursk 5 and 6 and Smolensk 4 in Russia, Chernobyl 5 and 6 in Ukraine and Ignalina 3 in Lithuania. At the Kursk RBMK plant, where Unit 5 was originally scheduled to come on line in 1995, an MKER-800--an "improved" graphite-moderated, channel-type reactor--is now planned for construction.

Since the Chernobyl accident, a considerable amount of work has been carried out--both by Russian institutions and by international groups--to improve RBMK reactor safety and to eliminate the root causes of the 1986 Chernobyl accident. Additional measures are planned or under way. But some concerns remain, particularly with respect to RBMKs of the first generation.

Principal Strengths:

• The low core power density of RBMKs provides a unique ability to withstand station blackout and loss of power events of up to an hour with no expected core damage.
• The units can be refueled while operating, permitting a high level of availability.
• The graphite moderator design allows the use of fuel that is not suitable for use in conventional water-moderated reactors.

Principal Deficiencies:

• The most significant difference between the RBMK design and most of the world's nuclear power plants is the RBMK's lack of a massive steel and/or concrete containment structure as the final barrier against large releases of radiation in an accident. The effectiveness of American-style reactor containments was shown in the 1979 Three Mile Island Unit 2 accident, when virtually all radiation was retained inside the containment building, despite considerable melting of the fuel. In the Chernobyl accident, the RBMK plant's accident localization system (the RBMK's version of containment) could not withstand the force of the accident. However, because the estimated energy released by the explosions was greater than most containment designs could withstand, it is highly unlikely that a containment structure could have prevented the release of radioactive material at Chernobyl.
• Accident mitigation systems are limited and ineffective.
• Reactor control systems are unforgiving to many potential system upsets, with a consequent potential difficulty of successful recovery.
• The reactor produces faster and less stable nuclear chain reactions--and power increases--when cooling water is lost. In technical terms, this characteristic is called a "positive void coefficient." Soviet engineers sought to mitigate this tendency by backfitting RBMKs with faster-acting control rods and other improvements. Modifications made to all RBMKs are generally considered to be adequate to maintain this positive void defect at a low enough level to preclude the type of nuclear excursion--a sudden, rapid rise in power level--that occurred at Unit 4. U.S.-style light water reactors are designed with just the opposite characteristic--a "negative void coefficient"--so that the nuclear chain reaction automatically stops when coolant is lost. The design of the Kursk 5 RBMK has reportedly been modified, resulting in a negative void coefficient.
• Inadequate fire protection systems.
• Limited capability for steam suppression in the graphite stack.
• Flawed separation and redundancy of electrical and safety systems.
• Complicated piping arrangements.

November 1995

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