Multiple fast reactors with various types of liquid-metal coolants are under development in Russia. The modular fast-reactor technology discussed here, which uses a lead–bismuth coolant, has been developed by the Institute for Physics and Power Engineering state research centre and JSC OKB Gidropress [1]. It relies on the unrivalled expertise in development and operation of lead–bismuth cooled Alpha-class Russian nuclear power submarines (15 reactors, 80 reactor-years), and in operational experience of sodium-cooled fast reactors.

Compared to sodium coolant, lead–bismuth coolant’s low heat-transfer properties tend to lower the power of the system of the core and slow the short plutonium doubling time, even with a conversion ratio much greater than unity. At the same time, natural properties of the heavy liquid metal coolant make it possible to simplify and markedly lower the cost of the reactor facility. For this reason, with this technology we can eliminate any conflict between economy and safety requirements.

This article elaborates the main features of the innovative concept of this modular fast reactor; substantiates the claimed safety level; analyses the possibility of using these reactors for multiple purposes, including reactors for export out of Russia designed for regional power production, duly considering non-proliferation requirements; and delineates the project status.

The SVBR 100 reactor system

The SVBR 100 reactor system (in English, the Lead-Bismuth Fast Reactor) has been designed to be an integral system with an electric power production of about 100 MWe for multi-purpose use within modular nuclear power plants or as self-contained energy sources for regional power production [2].

Basic features:

  • It incorporates a fast reactor with non-reactive lead–bismuth coolant (LBC), a eutectic lead–bismuth alloy in the primary circuit. Its melting point is 123.5°C and its boiling point is 1670°C.
  • The entire primary equipment circuit is contained within a robust single reactor vessel; LBC valves and pipelines are all exterior.
  • A protective enclosure surrounds the single-unit reactor vessel.
  • The reactor passes heat to a two-circuit heat-removal system and steam generator with multiple-circulation secondary coolant system.
  • Natural circulation of coolant in the reactor heat removal circuits is sufficient to passively cool down the reactor and prevent hazardous superheating of the core.
  • Compared to other reactors, the SVBR 100 reactor system has a dramatic reduction in the number of active safety systems; normal-operation systems ensure the performance of safety functions.
  • Main components in the single-unit reactor and reactor system are built as modules, and so can be replaced or maintained.
  • At the end of a core cycle, provision has been made for a single refuelling operation in which the entire fuel core is replaced as a single assembly.
  • Without design change, the reactor can use various types of fuel (uranium dioxide, MOX, nitride fuel). When using the latter two, the core conversion ratio is greater than zero, so it is possible to operate a self-supported closed fuel cycle.
  • Primary loop equipment repairs and fuel reloading can be performed without having to drain the LBC, whose liquid state is maintained by core decay heat and heating system operation.

The proposed reactor technology is based on the 40-year experience of lead–bismuth-cooled fast reactors in nuclear-powered submarines, and in ground-based bench prototypes. Numerous scientific and technical challenges have been overcome during development of this technology.

First was assurance of construction material corrosion stability, and operational monitoring and maintenance of coolant quality. Research has shown that reliable operation of the reactor system requires the measurement and control of only one parameter, the concentration of dissolved oxygen in the lead–bismuth coolant. This operation can be performed automatically.

Second was the major problem of dealing with radiation safety issues arising from Polonium-210 formation by irradiated bismuth. Staff (military and civilian) involved had periodic medical examinations. Analysis of multiple radiometric analyses of biological assays established that there were no cases of ingestion of polonium aerosol above permissible levels. The data demonstrates the effectiveness of the personal and collective radiation protection equipment, suitable technology choice and organisation [3].

The single-unit (integral) layout of the SVBR reactor system’s primary coolant equipment, and single-unit reactor vessel protective enclosure completely prevent coolant leaks. The probability of radioactive gas release has been lessened by pressurising the argon in the gas system to near atmospheric pressure.

A US study [4] includes a retrospective analysis of a cohort of about 4500 workers involved in handling Po-210 from 1944-1972. The sample was controlled with respect to internal exposure to Po-210. The authors concluded that there was no connection between the internal radiation dose from incorporated polonium up to 1 Sv (100 Rem) and the death rate due to malignant tumours. In fact, trends indicating mortality from cancerous diseases in the cohort were negative; the cohort was slightly better off than in reference groups with no polonium exposure.

Operational experience has shown that hardly any liquid radioactive waste is generated by the reactor system, since there is no need to decontaminate the primary coolant system.

The reactor system development project had a conservative approach to engineering design; the project has mainly included off-the-shelf or scaled engineering solutions with reasonable best practices proven by operational experience in nuclear power station reactors and other reactor facilities. This approach includes basic items, assemblies and pieces of reactor equipment: fuel pellets, fuel cladding, fuel assemblies, control rods, vessel internals, control rod actuators, LBC system components, steam generators with coaxial (fild) tubes, separators, self-contained cooldown condensers, gas system condensers, refuelling system equipment, and so on. The operating parameters of the primary and secondary cooling circuit, and the manufacturability of the design relative to the capabilities of industrial plants, were also developed conservatively.

Such an approach has considerably decreased technical and financial risks by reducing the likelihood of errors peculiar to innovative nuclear projects, and decreased the scope, terms and costs of R&D.

Passive safety

The combination of a fast reactor design with heavy metal coolant operating in an integral reactor layout ensures that the SVBR 100 reactor system meets IAEA international project safety levels for prevention of severe accidents and inherent safety, according to analysis and studies [5].

The reactor is typified by negative feedback and a negative void reactivity effect. Those tendencies, along with the control and protection system engineering design, prevents instantaneous neutron excursion.

The coolant’s high boiling point improves core heat removal reliability and safety since there is no departure from nucleate boiling. That property, plus the unit’s protective enclosure, prevents loss of coolant accidents (LOCA) and high-pressure radioactive release.

Low primary pressure reduces the risk of air leaks into the system, and allows a thinner reactor pressure vessel. Low operating pressure also lessens the limits on the rate of temperature changes under cycling conditions.

Nuclear steam supply systems are free of components that release hydrogen from thermal, radiation and chemical reactions with coolant, water and air. Therefore, a case of a primary-circuit depressurisation should not result in a fire or chemical explosion.

The design of coolant circulation inherently prevents water/steam ingress into the core in case of steam generator leaks.

Nuclear steam system supply (NSSS) safety is independent of the status of turbine-generator set systems and equipment, which may be designed and manufactured to general industry rules and regulations.

The existence of inherent NSSS safety features allow safety functions to be combined with normal-operation systems. Safety systems exclude components that could disable the system in cases of failure or wilful malicious action.

Decay heat removal (independent of the steam generators) occurs passively through the natural circulation of lead-bismuth coolant in the primary circuit by heat transfer from the single-unit reactor vessel to the Passive Heat Removal System (PHRS) tank water. It continues further by boiling tank water passing through the steam dump into the atmosphere. The reactor’s safe non-intervention period is about three days during which time no temperature limit is exceeded.

Steam generator leaks are confined passively, with the steam pressure in the gas system exceeding 1 MPa. If some steam generator tubes are ruptured, or if the gas system condenser breaks down, the burst rupture diaphragm is broken and leaked steam is dumped into the PHRS tank used as a bubbler (under normal circumstances, it serves a neutron protection function). Operational experience has found that a small SG leak need not cause an immediate NSSS shutdown.

Auxiliary protection system rods installed in dry channels are passively actuated by gravitational force. The rods, which have no drives on the reactor pressure vessel head, are held fixed in the raised position under normal temperature regimes by retainers made of an alloy with a particular melting point. When the lead–bismuth coolant temperature exceeds the specified value, it melts the retainers, and the rods drop.

Even with the combination of postulated initiating events such as damage to the containment, damage to the reinforced concrete slab above the reactor, depressurisation of the primary gas system with the direct contact of lead–bismuth coolant reflector in the single unit vessel with atmospheric air, and major plant blackout, there will be no reactor runaway, explosion, fire, nor environmental radioactive release exceeding the limits of NPP site population evacuation. Analysis suggests the probability of severe core damage is considerably lower than regulatory limits.

One of the main factors determining the high safety level of the reactor is the low potential energy stored in the coolant under operating parameters: 1 GJ/m3 for lead–bismuth coolant, versus 10 GJ/m3 and 20 GJ/m3 for sodium and water [6]. This feature explains the tolerance of the NSSS to not only equipment failures, human error or combinations of the two, but also to malicious actions in which all of the special safety systems are wilfully disabled.

Modularity

The ability to integrate the unit into a modular nuclear steam generating plant has multiple benefits.

A modular plant ensures a higher level of reliability and safety than a power plant with a single high-power reactor. It has higher reliability because the failure tolerance of the unit is a function of all the individual reactor systems. It has higher safety since there is lower potential radiation risk if one unit should encounter a problem.

A modular plant removes the need for a standby power system for regional nuclear power plants, where power supply is decentralised.

A modular plant provides for a load factor of not less than 90%, depending on the reliability performance of the turbine system, in cases of a long reactor operation period without refuelling. With regular refuelling and maintenance shutdowns, power decreases in the unit are much less than with a single, high-power reactor unit.

A modular plant can be manufactured in a large-scale (assembly-line) operation, on the order of 10 pieces per year, which considerably reduces manufacturing costs. Since operating specifications do not require the unique fabrication tools needed to make the high-pressure vessels of thermal nuclear reactors, a competitive market of multiple potential manufacturers could be established.

For this reason, standardised design methods can be developed for units of different capacities, and production line methods of construction and installation can be used. Combined with the highly serial manufacturability of the NSSS, these methods reduce the construction time and costs of these units to a level comparable with those of state-of-the-art steam and gas thermal (fossil fuel) power plants, although with a much lower cost per unit of electrical power generated than they can manage.

A modular plant with a lower mean power output can be installed in centres of power consumption, which prevents the need to construct high-capacity power transmission lines.

A modular plant can be commissioned in a stepwise fashion, with a phased power build-up as installation, startup and adjustment works proceed for a group of units. This structure reduces the return on investment time compared with a single higher-power reactor.

These benefits improve the attractiveness of the SVBR reactor system to customers. The shorter investment period of the reactor, as a product of the modular NSSS, and factory delivery of the assembled modules, is essential for its performance indicators to match up to thermal power plants.

Because the reactor has only two operating modes—enabled or disabled—one operator can control the NSSS of the plant. In case a unit fails, it is automatically taken out of operation and cooled down off-line.

At end of life (50-60 years), the spent fuel and lead–bismuth coolant are discharged. Then the main component of the unit, the single reactor unit, is swapped out with a new unit. The old unit may be dismantled and placed in a solid radioactive waste storage facility. Other components may be dismantled and replaced. The lifetime of the modular NPP will be limited only by the lifetime of the reinforced concrete civil engineering structures, and may reach 100-120 years. This long lifetime reduces the costs of construction/unit power output below that of nuclear new-build construction. When the time comes to finally decommission the NSSS building, there will be virtually no radioactive materials left after the unit is removed, which substantially reduces costs.

Multiple uses

The characteristics of the SVBR 100 reactor system suit its use for diverse applications as unified power plants with electric power output of about 1000 MWe.

Modular reactors could be used to build regional nuclear power plants and nuclear cogeneration plants with lower power located near towns. This application could be envisaged for developing countries without advanced electrical transmission and distribution technology, without access to fossil fuels, and without access to the large lump-sum financing required for higher-power nuclear reactors.

Modular reactors could be used to revamp nuclear power plants whose reactors have reached their end of life. According to this idea, the necessary number of SVBR reactor systems would be installed near the steam generators and reactor coolant pumps to generate an equivalent amount of steam to that which the previous reactor had generated. The results of an engineering capability and economics feasibility study show that such a revamp of Novovoronezh 2, 3 and 4 would cost half as much in unit capital cost as construction of a replacement facility. Such a revamp would also cut decommissioning costs dramatically. A long-term revamp programme would help maintain the viability of NPP satellite towns, as well as grid, transport and water infrastructure.

Modular reactors could supply power-intensive process facilities with electricity and steam at a stable price.

Modular reactors could be used as desalination plants or offshore NPPs. In this case, such NPPs could be procured using build-own-lease (or operate) contracts, which fulfil all the IAEA non-proliferation requirements.

Further development

The SVBR 100 reactor system is a first-generation design. As research and development activities progress and as operational experience accumulates, further improvements can be made. First, the use of a single-flow steam generator producing superheated steam will improve the efficiency of the thermodynamic cycle, lower capital costs and simplify the design of the reactor system. Second, a higher reactor outlet temperature, in which the maximum fuel cladding temperature would rise from 600°C to 650°C, would increase reactor thermal power by 10% without changing its design or cost. Third, the use of nitride fuel could reduce fuel costs and increase reactor fuel cycle length from seven years for uranium dioxide to 15 years (with fuel element operability confirmed).

Project status

The operating conditions of lead–bismuth cooled reactors in nuclear power submarines differ markedly from their planned use in nuclear power plants. While the submarine power reactors mainly operated at lower power levels with low lead–bismuth coolant temperatures. The NPP reactor facility would run primarily at the rated power level. Also, the life expectancy requirements of the NPP system components are much greater than those of the submarine system components.

For these reasons, a trial commercial SVBR 100 reactor system unit needs to be built. The construction costs of this prototype are non-recurrent, since nuclear units of different capacity and for different purposes may be made on the basis of the standardised reactor module as built, with no additional R&D, as discussed above.

The NPP prototype will be equipped with additional transducers and devices. Once built, it will be possible to demonstrate its passive safety features under controlled conditions by creating any combination of equipment failures, human errors and simulation of wilful malicious actions.

Once certification tests of the prototype are complete and design features are confirmed, the SVBR reactor system can be prepared for commercialisation.

On 15 June 2006, Rosatom scientific and engineering council no. 1 recommended that a detailed design of a SVBR 100 reactor trial unit would be pursued with reference to a specific site. The project is managed by AKME-Engineering, a 50-50 joint venture between state corporation Rosatom and JSC EuroSibEnergo.

As of May 2011, siting licence works are underway, pilot plant specifications and key reactor and reactor core research and development works have begun. A complete reactor and power plant design is expected to be completed by 2013, along with a preliminary safety report. In 2013 a construction licence is also expected to be obtained.

The trial unit is expected to be commissioned by 2017 at the Russian state Atomic Reactor Research Institute, Dimitrovgrad, in the region of Ulyanovsk. The total investment cost is estimated at RUR 16-18 billion ($500-600 million).

 

This article was originally published in the October 2011 issue of Nuclear Engineering International (p20-25)

By V. V. Petrochenko, Director General JSC AKME-engineering, 24 Bolshaya Ordynka Street, Moscow 119017, Russia.

 

References

1. Toshinsky G.I., Stepanov V.S., Petrochenko V.V. et al. "SVBR-100 module-type fast reactor of the IV Generation for regional power industry." International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, IAEA-CN-176-FR09P1132, 7-11 December 2009 Kyoto, Japan.

2. Zrodnikov A.V., Toshinsky G.I., Stepanov V.S. et al. "Innovative nuclear technology based on modular multi-purpose lead-bismuth cooled fast reactors." Progress In Nuclear Energy, Vol. 50, pp.170-178, 2008.

3. Zrodnikov A.V., Toshinsky G.I., Dragunov Yu.G., Stepanov V.S.et al. "Nuclear power development in market conditions with use of multi-purpose modular fast reactors SVBR-75/100", 2006, Nuclear Engineering and Design, Vol. 236, pp. 1490-1502.

4. Laurie D. Wiggs, Carol A. Cox-De Vore and George L. Voelz. Mortality among a Cohort of Workers Monitored for Po-210 Exposure: 1944-1972 y.y. Epidemiology Section Occupational Medicine Group. Los-Alamos National Laboratory. Health physics, Vol. 61, No 1, 1991.

5. Toshinsky G. I., Komlev O.G, Stepanov V.S. et al., "Principles of Providing Inherent Self-Protection and Passive Safety Characteristics of the SVBR-75/100 Type Modular Reactor Installation for Nuclear Power Plants of Different Capacity and Purpose", Paper No. 175598. International Conference Advanced Nuclear Fuel Cycles and Systems (Global07), September 9-13, 2007, Boise, Idaho, USA.

6. Toshinsky G.I, Komlev O.G, Tormyshev I.V. et al. "Effect of Potential Energy Stored in Reactor Facility Coolant on NPP Safety and Economic Parameters". Proceedings of ICAPP 2011, Nice, France, May 2-5, 2011, Paper 11465.