18.3.1 Introduction of ACP100

Since 2010, one type of small and medium-size water-cooled pressurized reactor called ACP100 has been developed by China’s national nuclear corporation. ACP100 is an innovative reactor based on existing PWR technology, adapting a ‘passive’ safety system and ‘integrated’ reactor design technology. After 3 years’ development, its overall design was complete, including conceptual design and basic design. A number of testing facilities are under construction and R&D on safety-related experiments will be carried out in the following 2 years. The construction of the ACP100 will be started around the end of 2014.

18.3.2 Technical aspects

The ACP100 design has the following remarkable technical features:

• Primary system and equipment integrated layout. The maximum size of the connection pipe is 5-8 cm, whereas in a large NPP it is 80-90 cm.

• Large primary coolant inventory.

• Small radioactivity storage quantity. The total radioactivity of an SMR is one-tenth that of large NPPs, and meanwhile a multi-layer barrier is added to keep the accident source-term at a low level.

• The vessel and equipment layout is beneficial for natural circulation.

• The decay heat removal is more effective: two to four times the efficiency of large NPP heat removal from the vessel surface.

• Smaller decay thermal power: one-fifth to one-tenth times of decay thermal power compared with that of a large PWR after shutdown, and is easier to achieve safety by way of the ‘passive’ system.

• Reactor and spent fuel pool are under ground for better protection against exterior accident and for the reduction of radioactive material release.

• No operator intervention is needed for 72 hours after an accident.

• Passive severe accident prevention and mitigation action, such as for containment hydrogen eliminator, cavity flooding, etc. is built in to ensure the integrity of pressure containment.

• The modular design technique makes it easy to control the product quality and shorten the site construction period.

The DMS is a typical simplified BWR and was proposed by Hitachi, Ltd (Kawabata et al., 2008) in cooperation with the JAPC. The targets for the design are the same as in the IMR and are summarized in Table 19.1. A schematic view of the reactor is shown in Figure 19.4 and its major specifications are summarized in Table 19.4.

Containment

vessel

Reactor

vessel

Main

steam

line

The core is shortened and cooled by the natural circulation of the coolant without the recirculation pumps. One of the characteristic features for the system simplification for this concept is elimination of the steam-water separator and dryer, by introducing a gravity separation mechanism for the steam-water separation process. Therefore, the reactor pressure vessel is simplified and compact.

The power output is 428 MWe. The average discharge burnup is 45 GW d/t and the refueling interval is 24 months. The effective length of the core is shortened to 2 m to increase the driving force for the natural circulation core cooling. The system pressure is 7.2 MPa and is the same pressure as in the normal BWR. For the safety system design, the DMS employs a simplified ECCS with passive accumulator system, and hence, eliminated the high pressure core flooder (HPCF) system and decreased about 60% of the reactor core isolation cooling (RCIC) system capacity.

V. Kuznetsov Consultant, Austria

17.1 Introduction

The category of small modular reactors (SMRs), includes those with an equivalent electric output less than ~300 MW(e), having a high degree of factory fabrication allowing for transportation of factory-assembled reactor modules or even the whole plant by barge, rail or truck, and with an option to build power stations of flexible capacity through a multi-module approach. Such designs are being developed in the Russian Federation with the view of providing a secure energy supply to small regional energy systems located in remote and hard to access areas of the country where the climate is characterized by extremes and the transportation routes providing connections to the rest of the country are unreliable and available only over a short season during the year. The overall Russian strategy is to have reactors of different capacities from large and very large (1200-1500 MW(e)) to small and very small (10-150 MW(e)), including medium-sized (300-600 MW(e)), to cater for a variety of centralized and regional energy needs stemming from the geographical and climatic conditions of the country (Velikhov, 2008).

Some 60-70% of Russian territory is affected by permafrost which complicates large-scale construction and makes it very expensive to develop and maintain reliable transportation routes. These are the large territories in the north and east of the country, which are characterized by sparse population concentrated mostly around mining and raw-material reprocessing enterprises and military bases. The temperatures in winter may be extremely low and heat for residential needs is, therefore, in demand as well as electricity. Connections to the more populated areas of the country are seasonal (not available in winter) and unreliable (may be not available in some years due to the damage caused by the permafrost melting). In such an environment basic requirements to an energy source are the ability to operate over long periods without the need of fuel delivery and the ability to operate in a co-generation mode, producing heat as well as electricity.

The territories in the Russian north are rich in oil, gas, alumina, nickel, diamonds and other valuable natural resources, the development of which is crucial for the still largely resource export-oriented Russian economy. One of the factors associated with mining is the lifespan of a mine which is often limited to just a few decades (IAEA, 2007). In view of this, relocatable energy sources may have an advantage.

In its north and east, Russia has a long coastal line, with the seaside being covered with ice over a long winter. Small and sparse settlements located along or nearby this

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coastal line (typically, they are settlements around local small enterprises or military bases) also need energy and heat and may benefit from autonomous reliable energy sources matching their small or not present electricity grids. Here, transportable barge-mounted nuclear co-generation plants with the on-board fresh fuel reserves and spent fuel storage could be considered as candidates (IAEA, 2009).

Finally, the Russian Federation is considering a project to develop gas and oil production from the shelf of the Barents Sea, located in the country’s north. Nuclear reactors may be employed for such a production, including those located ashore, or on the sea bottom to power the underwater mining plant, or for propulsion purposes on submarines delivering gas to the on-shore terminal (Velikhov, 2008).

Based on data from Russian Federal Tariff Service (2013), the maximum electricity tariff observed in some areas in Russia’s north and east is 20 times higher compared to the minimum one (97 USD cents per kW h versus 5.0 USD cents per kW h). High tariffs mean lack of the centralized grids, difficult conditions for fuel delivery, limited demand or other specific siting conditions making it impractical or impossible to build an economy of scale large power plant. Such areas are, therefore, potential markets for SMRs.

Design development activities for civil SMRs started in Russia in the 1980s and borrowed extensively from the experience in design and operation of the marine propulsion reactors for the navy and the nuclear ice-breaker fleet. This positive experience includes several different designs of pressurized water reactors (PWRs) and one reactor cooled by lead-bismuth eutectics; it amounts to not less than 6500 reactor-years overall with over 260 reactor-years for the reactors of nuclear ice­breakers alone (Sozonyuk, 2011). For comparison, experience in the design and operation of conventional land-based PWRs — the reactors most commonly deployed worldwide, including those of the Russian VVER type — constitutes 8000 reactor — years (Sozonyuk, 2011).

As it will be shown in the following sections, the Russian SMR designs based on PWR technology incorporate both, proven features of the previous-generation marine propulsion reactors and the state-of-the-art features of contemporary VVER-type reactors. The SMR design employing lead-bismuth eutectics technology borrows both, from submarine reactors of the 1980s and from the experience of the Russian sodium-cooled fast reactors.

Comparing the development of SMRs in Russia with that in the USA, the following can be noted:

• The term ‘small modular reactors’, introduced in the US programes on small reactor development in the mid-2000s, is not in common use in Russia, although many of the Russian designs share common design approaches with the SMRs being currently developed in the USA. In this, Russian activities for such SMRs were started earlier, in mid-1980s.

• Unlike US designs, multi-module plants of flexible capacity and underground location of the reactor modules are not commonly considered in Russia. However, twin units are commonly considered and several of the Russian SMRs are designed to be located on non-self-propelled barges, while at least one design concept is being considered for seabed location.

• Russian designers commonly accept SMRs to be more expensive sources of electricity compared to the state-of-the-art large nuclear power plants (NPPs) and do not believe they may directly compete with larger plants, say, through shorter construction periods or multi-module plant configurations. Instead, they target particular niche markets where electricity costs are high, where co-generation, long refueling interval or plant relocatability are assets, where transportation routes are seasonal, where the demand is limited and siting conditions are specific (i. e., no water in winter due to deep freezing of rivers or other water reservoirs). According to Russian Federal Tariff Service (2013), there are many such niche markets in Russia and, based on data from p. 113 of ‘ Current status, technical feasibility and economics of small nuclear reactors’ (NEA-OECD, 2011), similar conditions are also being observed in several other countries. The above-mentioned unique standpoint is because Russian policy is to have large, medium and small reactors on a complementary, not a competitive, basis. Complementarity is pursued in view of different niches for reactors of different capacity available domestically and, potentially, worldwide.

• As in the US case, Russian SMRs are being designed and licensed to operate first in their country of origin. Should the experience of their operation be positive, they could later be offered on world markets with some features tailored for the needs of such markets, e. g., co-generation option with heat production changed to co-generation option with seawater desalination.

The objective of this chapter is to present the design and safety features, including defense-in-depth, probabilistic safety goals and, specifically, design features for protection against external event impacts, the design and operating characteristics and the anticipated applications for a variety of SMRs being developed or deployed in the Russian Federation.

In line with this objective, Sections 17.2-17.4 present SMR designs being developed or already developed by the Russian design organizations/companies OKBM Afrikantov (2013), AKME Engineering (2013) and NIKIET (2013), respectively. These sections also highlight the on-going R&D for the corresponding SMR projects.

Another objective of this chapter is to present the deployment status and prospects for the Russian SMRs, which is accomplished in Section 17.5. In this section the design and licensing status of SMRs is highlighted. Section 17.5 also provides the available economic data and presents the prospects for SMR deployment in the indigenous and foreign markets.

Section 17.6 presents the future trends for the barge-mounted, nuclear ice-breaker and land-based SMRs developed in the Russian Federation. A conclusion is drawn in Section 17.7 and, finally, section 17.8 provides a short description of sources of further information and advice, including a brief commentary on key publications and databases of International Organizations and national trade/professional bodies, research and interest groups and web sites. The chapter is concluded by a comprehensive list of references.

The data of Russian Federal Tariff Service (2013) indicates that the spread of electricity costs in the Russian Federation is quite even, leaving a potential market space not only for large and small but also for medium-sized reactors. Design concepts of such reactors are also being developed in the Russian Federation and, although they are beyond the scope of this book, Section 17.8 lists such design concepts and points to the best available sources of further information.

Half of the world’s population lives in cities, and this will grow as increased urbanization occurs in developing countries. Lagos, Dhaka, Manila and Cairo are examples of megacities (those with over 10 million inhabitants) in the developing world.

Air pollution and climate impacts of the world’s megacities are worsening (WMO/ IGAC, 2012; Cossardeux 2013). SMRs could be a source of emissions-free electricity to help power megacities and mitigate their effect on the air and the climate. To serve economic and city planning, they could be deployed as scalable multi-module plants ramping up to meet demand. In addition, they could be sited closer to load centres if a reduced surrounding safety zone due to their designs’ safety benefits (IEAE INPRO DF 5, 2012) could be envisaged.

The safety design philosophy of ACP100 is to realize a high level of safety and, at the same time, to simplify the design of the systems by means of a passive engineering safety system. No emergency diesel generator is needed. The emergency measures outside the plant boundary should be made technically unnecessary or reduced to a minimum. The passive system of ACP100 include: long-term residual heat removal system to cope with station blackout accident, passive core cooling system to cope with LOCA, cavity flooding during a severe accident, and a passive containment heat-removal system. The passive containment heat-removal system means that the heat is taken away by the gas and steam convection between containment and ultimate heat sink by natural circulation. This ensures the containment integrity under accident conditions.

• The absence of any large-diameter piping associated with the primary system removes the possibility of large-break LOCAs. The elimination of large-break LOCAs substantially reduces the necessity for emergency core cooling system components, alternate current (AC) supply systems, etc.

• The large coolant inventory in the primary circuit results in large thermal inertia and a long response time in the case of transients or accidents.

• Inherent safety features include: an integrated primary coolant system, eliminating large break LOCAs; long characteristic times in the event of a transient or severe accident, due to large coolant inventory and the use of passive safety systems; negative reactivity effects and coefficients.

• Passive safety systems are duplicated to fulfil redundancy criteria. According to Chinese nuclear safety regulations, the shutdown system is diversified. The systems include a residual heat removal system, an emergency injection system, and safety relief valves which protect the reactor pressure vessel against over-pressurization in the case of strong differences between core power and the power removed from the reactor pressure vessel. All safety systems mentioned in this paragraph are passive systems.

The GTHTR300 is a high-temperature helium gas-cooled modular reactor with a high thermal efficiency gas turbine cycle system, and was proposed by the Japan Atomic Energy Agency (JAEA) (Kunitomi et al., 2002, 2004). The targets for the design are summarized in Table 19.5 as the user requirements. A schematic view of the reactor concept is shown in Figure 19.5 and its major specifications are summarized in Table 19.6. The GTHTR300 is characterized by the significantly simplified system design, the high thermal efficiency and the improved economy from utilizing the proven technologies obtained with the JAEA’s HTTR (high temperature engineering test reactor), as well as the good safety features.

The power output is 280 MWe. The average discharge burnup is 120 GW d/t and the refueling interval is 2 years. The fuel is the block type one with the average U enrichment of 14%. The core is the pin-in-block and annular type, and its effective height is 8 m. The core is cooled by the helium gas and the core inlet/outlet temperatures

Recuperator

Precooler

Turbine Generator

Compressor

Figure 19.5 Schematic view of GTHTR300 reactor concept (taken from Kunitomi et al., 2002).

are 860/1123 K under the system pressure of 6.8 MPa. For the safety system design, the GTHTR300 concept employs the target of ‘severe accident-free’ (Katanishi et al., 2003). This can be attained by its special characteristics as follows: there is no possibility for core melt and the core can stand high temperatures, because the fuel is ceramic and the graphite materials and the core structures are graphite. Even if abnormal events occur, the transients are expected to be slow due to the low core power density and large core heat capacity as well as the negative reactivity feedback. In addition, the coolant, i. e. the helium gas, is very stable and does not result in any mechanical energy production or release. The maximum fuel temperature can be lower than 1873 K (1600 °C) even in an accident, and hence the integrity of the fuel, the coated fuel particle, is assured. Therefore, containment is not necessary by adopting the confinement, which is the reactor building with a low leak rate less than about 1%/d.

The basic design was finished by the end of the 2003 fiscal year and was reviewed by electric companies, vendors and universities. According to the preliminary cost evaluation results, the construction cost was about 200 000 yen/kWe and the electricity generation cost was about 4.5 yen/kW h. The design of the co-generation system of the GTHTR300C was also performed for hydrogen production.

After that, based on this concept, the basic design of a small reactor concept of HTR50S of 50 MWt output is underway (Ohashi et al., 2011) in collaboration with Kazakhstan. This concept utilizes a steam generator system under a reactor outlet temperature of 1023 K (750 °C), and is intended to be realized in the 2020s. Also, the MHR-50 (Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor — 50) concept has been developed by MHI (Shimizu et al., 2011) in cooperation with the JAEA. This is a high temperature helium gas-cooled modular reactor concept of 50 MWe output with a steam generator and a steam turbine cycle system, utilizing technologies verified by the JAEA’s HTTR as much as possible.

The HTTR is a high-temperature helium gas-cooled test reactor with an output of 30 MWt, constructed to establish and advance the technical bases of the high — temperature gas-cooled reactor and irradiation tests in the high-temperature core (Saito et al., 1994). The first criticality was achieved in November 1998, and full power was reached with the core outlet coolant temperature of 1123 K (850 °C) in December 2001. The maximum coolant temperature of 1223 K (950 °C) was achieved in April 2004. The core inlet coolant temperature is 668 K (395 °C). The major specifications of the HTTR are summarized in Table 19.7.

At present, the reactor has been utilized for safety demonstration tests (Nakagawa et al., 2004). These tests are performed to demonstrate its inherent safety characteristics under international collaboration. Tests such as for control rod withdrawal, loss of core flow and loss of core cooling are included. Also, to establish the technical bases for the nuclear heat utilization, related R&D for the IS (iodine-sulfur) process to produce hydrogen from the water has been performed.

Small modular reactors that are being developed or had been developed by the Russian ‘Afrikantov Experimental Design Bureau for Mechanical Engineering’ (OKBM Afrikantov) include:

• KLT-40S reactor for the twin unit barge-mounted (floating) NPP with a co-generation option;

• ABV reactor for the twin unit barge-mounted or land-based NPP with a co-generation option;

• RITM-200 reactor for a new generation multi-purpose nuclear ice-breaker, with a potential of being considered further for the barge-mounted and land-based NPPs.

All of the above-mentioned designs are indirect cycle PWRs. In most cases twin units with such reactors are being considered as standard plant configurations. Multi-module plant configurations are not being considered for any of the OKBM Afrikantov designs. The design and operating characteristics of these SMRs are summarized in Table 17.1, and the core and fuel design characteristics in Table 17.2.

The most mature of the SMR designs by OKBM Afrikantov is the KLT-40S reactor of 38.5 MW(e) per module. It has a submarine prototype experience of several thousands of reactor-years and an ice-breaker prototype experience of nearly 300 reactor-years. The KLT-40S design appears to be similar to a conventional PWR reactor with separate vessels accommodating the reactor core and internals, the main circulation pumps, the pressurizer and the steam generators (see Figure 17.1). Control rod drives are external and located above the reactor vessel lid. However, in the KLT-40S case all separate vessels are compact and connected by short pipelines with vessel penetrations in the hot legs. The piping is double wall and incorporates so-called narrowing devices near the penetrations. All modules are connected with brackets fixed by bolts and nuts to restrict possible leaks in the case of a pipe rupture (not shown in Figure 17.1). All primary coolant systems are located within the primary pressure boundary. The above mentioned features effectively reduce the scope and magnitude of possible loss-of-coolant accidents (LOCAs). The refueling is performed in batches every 45-46 months.

The ABV design shown in Figure 17.2 is also backed by some prototype operation experience, the details of which have not been disclosed. It incorporates an integral primary circuit design with in-vessel steam generators. However, the pressurizer and the control rod drives are external. The core design and dimensions are similar to those of the KLT-40S. The reactor is refueled at the factory each 12 years (whole core refueling). To support achieving a long refueling interval, chromium-nickel alloy is used as cladding material instead of the traditional zirconium alloy. The newer version of the ABV reactor employs pumps for the primary circulation, while the older version relied on natural circulation of the primary circuit (ARIS, 2013). The RITM-200 — the newest of all SMR designs presented in this section — has a

per-module electrical output of 50 MW(e). It incorporates the experience in design and operation of many of the Russian marine propulsion reactors. Like ABV, RITM-200 has an integral primary circuit design with in-vessel steam generators (see Figure 17.3). The pressurizer and the control rod drives are again external. The core design and dimensions are modified against those of the KLT-40S (see Table 17.2), but exact data on these modifications is not available. The reactor employs forced circulation of the primary coolant provided by the main circulation pumps, as shown in Figure 17.3. The reactor is refueled at the refueling base each 4.5-7 years (whole core refueling). Chromium-nickel alloy is used as cladding material instead of the traditional zirconium alloy (Veshnyakov, 2011).

The SMR designs from OKBM Afrikantov have the following common design features:

• All reactors are being designed for twin unit plant configurations with individual turbine generators in the power circuits of each of the reactors.

• Flexible co-generation options (production of heat for residential heating or superheated process steam or seawater desalination) are being provided for in the designs as options.

• Construction period is typically 4 years (perhaps, somewhat shorter for the newest RITM — 200).

• All designs provide for load following capability.

• The fuel is UO2 in inert matrix within the cylindrical fuel elements and the fuel enrichment is always below 20%, even for the ice-breaker reactor RITM-200.

Exchanger of i circuits

Hydraulic

accumulator

Figure 17.1 Layout of the KLT-40S reactor (ARIS, 2013).

• All designs, including those with the integral primary circuit, have external control rod drives, external pumps and external pressurizers which use gas as a working medium to ensure a slow response of the primary pressure to rapid changes of the core temperature.

• Mechanical control rods are used for reactivity control in all designs and, additionally, liquid boron system is used in the ABV reactor only.

• The average fuel burn-up is within the range of 45-49 MWday/kg, i. e., below the values achieved in the state-of-the-art NPPs with large reactors.

• All designs employ double containments.

The major differences between the designs are as follows:

• The KLT-40S employs an older compact modular design of the primary circuit, while the ABV and the RITM-200 employ integral designs of the primary circuit with in-build steam generators.

• The KLT-40S is refueled in a conventional batch mode each four years while the ABV and the RITM-200 provide for factory refueling with a longer interval between refueling of up to 7-12 years.

Reactor

Figure 17.2 Layout of the ABV reactor, reproduced with permission by the IAEA from reference IAEA (2007).

• Chromium-nickel alloy is used as cladding material in the ABV and the RITM-200 to ensure operation with a long interval between refuelings.

• Liquid boron system is used for reactivity control in the ABV design, while no such system is being provided for in the KLT-40S and RITM-200 designs.

Comparison of the SMR designs from OKBM Afrikantov indicates the designers prefer to follow an evolutionary approach within which new components are introduced with caution and that well-proven components are used in new designs whenever applicable. The major new direction is related to the integral design of the primary circuit within which only steam generators are being placed inside the reactor vessel. The integral design makes it possible to increase electric output (compare KLT-40S and RITM-200 in Table 17.1) and reduces the dimensions and the mass of a nuclear island (for a twin unit plant the corresponding reduction in mass is from 1770 t (KLT-40S) to 1100 t (RITM-200)). The core design of the RITM-200 also makes it possible to increase by a factor of 20 the number of cycles of power change allowed during the core’s lifetime (Veshnyakov, 2011).

The overall approach to safety design is similar to that of the state-of-the-art PWRs, with some features borrowed from the experience of the marine propulsion reactors. The designers have attempted to eliminate or de-rate as many accidents as possible ‘by design’ and then dealt with the remaining ones by plausible combinations of the redundant and diverse active and passive systems (IAEA, 2009). A short summary of the safety design features incorporated in the SMRs from OKBM Afrikantov is provided below, in line with the defence in depth strategy defined in the Safety Standard No. SSR-2/1 Safety of Nuclear Power Plants: Design, Specific

Main circulation pump

Safety Requirements (IAEA, 2012a). The focus is on design features specific for the addressed SMRs.

According to IAEA (2012a), cited with permission by the IAEA, ‘the purpose of the first level of defence is to prevent deviations from normal operation and the failure of items important to safety’. For this level the OKBM designs provide ‘sound and conservative designs, siting, construction, operation and maintenance’ and, specifically (IAEA, 2007, 2009; ARIS, 2013; Veshnyakov, 2011):

• negative reactivity coefficients over the whole operation cycle;

• compact modular designs with no long pipelines (KLT-40S) or integral primary circuit designs with minimized reactor pressure vessel penetrations (ABV, RITM-200), to eliminate or de-rate certain groups of LOCA;

• eliminated liquid boron system (KLT-40S, RITM-200).

‘The purpose of the second level of defence is to detect and control deviations from normal operational states in order to prevent anticipated operational occurrences at the plant from escalating to accident conditions’ (IAEA, 2012a). Specific systems and features are provided in the OKBM Afrikantov designs to support meeting the objective of Level 2, including (IAEA, 2007, 2009; ARIS, 2013; Veshnyakov, 2011):

• state-of-the-art I&C systems based on PWR and marine propulsion reactor experience;

• redundant and diverse reactor shutdown systems (e. g., control rods driven by gravity, force of springs and mechanically);

• relatively large coolant inventory and high heat capacity of the primary circuit or nuclear installation as a whole, resulting in larger thermal inertia of the system;

• external pressurizers operated on gas medium to ensure slow primary pressure increase in transients.

‘For the third level of defence, it is assumed that, although very unlikely, the escalation of certain anticipated operational occurrences or postulated initiating events might not be controlled at a preceding level and that an accident could develop. In the design of the plant, such accidents are postulated to occur. This leads to a requirement that inherent and/or engineered safety features, safety systems and procedures be provided that are capable of preventing damage to the reactor core or significant off-site releases and returning the plant to a safe state’ (IAEA, 2012a).

To meet the objective of Level 3 the OKBM designs include redundant and diverse active and passive reactor shut down, emergency core cooling and residual heat removal systems (ARIS, 2013; IAEA, 2007, 2009; Veshnyakov, 2011). In this, they are more on the side of the state-of-the art Russian VVER type reactors, as passive decay heat removal systems are seldom found in the designs of the previous-generation marine propulsion reactors.

‘The purpose of the fourth level of defence is to mitigate the consequences of accidents that result from failure of the third level of defence in depth. The most important objective for this level is to ensure the confinement function’ (IAEA, 2012a). To meet this objective, all of the OKBM Afrikantov SMR designs provide for in-vessel retention of corium without a core catcher. The following features support achieving this quality:

• relatively low core power density (all designs omit KLT-40S) (see Table 17.1)

• active and passive systems of reactor vessel cooling.

In addition to this, all designs incorporate double containments (see Table 17.1) and active and passive containment cooling systems.

Finally, ‘the purpose of the fifth and final level of defence is to mitigate the radiological consequences of radioactive releases that could potentially result from accident conditions’ (IAEA, 2012a). To meet this objective, all OKBM Afrikantov SMR designs provide for on-site and off-site emergency measures and off-site emergency planning zones. However, the requirements to off-site emergency planning are reduced compared with those in place for conventional NPPs with large reactors. As an example, cited with permission by the IAEA from IAEA (2012a), the on-site and off-site emergency planning measures for a barge-mounted plant with the two KLT-40S reactors are as follows:

• Staff presence in the compartments adjacent to the containment and in other compartments with high radiation level to be excluded;

• To limit radiation dose to the population living within a 1 km radius from the floating NPP, depending on the actual radiation situation, some protection measures such as iodine prophylaxis or sheltering will be implemented;

• Temporary limitation will be established on the consumption of some agricultural products grown within a radius of 0.5 km from the ‘floating’ plant, when contaminated with radioactive release;

• Evacuation of the population is not required at any distance from the ‘floating’ NPP.

The SMR designs from OKBM Afrikantov incorporate protection against external event impacts. The plants are being designed for the design earthquake frequency 10-2/ year and the maximum design earthquake frequency 10-4/year, with maximum design earthquake 8 on the MSK scale (pre-Fukushima Daiichi data) (IAEA, 2009). In this, the equipment, machinery, systems important for safety and their mounting are designed for 3g Peak Ground Acceleration (PGA). They remain operable under inclination and heaving, typical of a floating NPP operation conditions (IAEA, 2009).

In addition to the above-mentioned, the designs of the barge-mounted plants take into account the external events specific of a barge-mounted NPP, such as the following (cited with permission by the IAEA from IAEA, 2009):

• Sinking of the floating NPP;

• Grounding, including that on a rocky ground;

• Collisions with other ships;

• Fall of a military aircraft onto the plant from a high altitude;

• Blockage of water intakes by debris from another ship;

• Helicopter crash-landing on a floating NPP.

Except for a nuclear ice-breaker plant with the two RITM-200 reactors, barge — mounted NPPs are being designed to produce energy only when rigidly moored to a shore (or a shore fixed structure) (IAEA, 2009). Plant mooring devices provide for plant retention at a tsunami wave height of up to 4 m. Barge-mounted plants are being designed to be moored in a bay protected by a dam (Sozonyuk, 2011) (see Figure 17.4).

The probabilistic safety analysis performed for the KLT-40S and ABV-based barge-mounted plants indicates the core damage frequency (CDF) is 10-6/year, while the large early release frequency (LERF) is 10-7/year (ARIS, 2013). These numbers are similar to the values typical of the state-of-the-art large NPPs with large reactors (ARIS, 2013).

In 2013, the pilot barge-mounted NPP with the two KLT-40S reactors, named ‘Akademik Lomonosov’, was under the final stages of construction in St Peterburg, Russia, with no additional R&D for this project being required. Subject to the results of pilot plant operation, the design features and main technical solutions developed for the barge are planned to be used in full in new generations of ‘floating’ NPPs,

equipped with nuclear reactors of other types. Specifically, it is planned to be done for the barge-mounted plant with the two ABV reactors.

The previous design of the barge-mounted plant with the two ABV reactors had already been licensed for construction and operation, but requests of customers in the Russian north indicate that a longer core lifetime is preferable in targeted locations. Negotiations with potential customers are in progress to tailor plant characteristics to their requests. Specifically, R&D is on-going to develop and validate the ABV core design with 12 year refueling interval with the initial enrichment of fuel by 235U not exceeding 20%.

For the new RITM-200 ice-breaker reactor, design has been completed and approved in March 2013, followed by the bidding which has been won in April 2013 by the ZiO-Podolsk Plant of the joint stock company (JSC) Atomenergomash. ZiO Podolsk will construct the nuclear installation, while the rectors will be assembled at OKBM Afrikantov. No additional R&D is needed for this design.

These might be markedly different from those in developed countries (Hecht, 2012). This has to do with differing value systems and weight given to different socioeconomic aspects, for example, antipathy to coal mining or conversely national reliance on coal mining; environmental desiderata; water scarcity; the existence of mining operations; the relative openness and transparency of energy policymaking and bid procedures; the potential for participation in the supply chain and for technology transfer. The pressing need for clean, sustainable power and water supplies in deprived and burgeoning countries alike might create greater openness for small-scale nuclear power in the form of SMRs. Moreover, public-acceptance factors are as potent or determinant for developed countries as for developing countries when the prospect of wide deployment of nuclear is raised, whether it be in the shape of SMRs or not. Fear of increased access of developing countries to nuclear power would contribute to a policy that locks countries with highest growth prospects and most acute need for low-carbon electricity into unsustainable, high-emitting, or costly renewable technologies (Ropeik, 2012). Nevertheless, how nuclear is construed is historically or culturally determined, and can change (Weart, 2012).

SMR provides an opportunity and vehicle for development and should not just be considered a tool, or even a goal in itself, to demonstrate technical advancement.

20.2 SMR choices in developing countries

This chapter posits SMR as a counterfactual choice — ‘what one would have chosen if one had the choice’ — for energy for developing countries (Sen, 1995). At an historical moment when developing countries disproportionately suffer the effects of climate change and, for some, the energy-security challenge of importing fuels with high price volatility, there is considerable value in the potential for an emissions-free energy source with comparatively stable fuel prices.

Embedded risk biases may currently prevent recognition of it; however, in the looming crisis of climate change, the true costs of business as usual will fall disproportionately on developing countries. That they may prove higher than going along the SMR route is a matter of opinion. However, it is not an outlandish conclusion to draw: ‘The biggest factor in the decision to construct a plant, may shift from the customer’s ability to finance the project to a careful consideration of opportunity cost,’ observe Abdulla et al. (2013), in the context of the possibility of more stringent carbon regulation.

As a developing country, if SMR were not available, what would be the alternatives? What are the potential consequences (including for the climate, societal development, the economy, and thus the ethical dimension), for developing and developed countries, viewed as a dyad?

18.3.7.1 Level 1: Prevention of abnormal operation and failure

Contributions of ACP100 inherent and passive safety features at this level are as follows: owing to the absence of large-diameter piping in the primary system, large- break LOCAs are eliminated; canned pump eliminates boron injection for pump sealing system.

18.3.7.2 Level 2: Control of abnormal operation and detection of failure

The ACP100 passive safety feature for this level is as follows: a large coolant inventory in the primary circuit results in a larger thermal inertia and in longer response times in the case of transients or accidents.

18.3.7.3 Level 3: Control of accidents within the design basis

ACP100’s safety systems are based on passive features obviating the need for actions related to accident management over a long period. Long-term residual heat removal systems cope with station blackout accidents, passive core cooling systems cope with loss of coolant accidents and cavity flooding during serve accidents, and passive containment of heat removal system.

The 4S concept is a small modular SFR with the good safety features and a very long operation period without refueling. It has been proposed by the Toshiba Corporation (Tsuboi et al., 2009, 2012) in cooperation with CRIEPI (Central Research Institute of Power Industries). The features of the 4S are summarized in Table 19.8. A schematic view of the reactor concept is shown in Figure 19.6 and its major specifications are summarized in Table 19.9. The 4S concept is characterized by the significantly simplified system design with nonrefueling, low maintenance requirements and high safety features; it is suitable for supplying energy to remote communities, mining sites, and so on.

Table 19.8 Features of a 4S reactor

Sodium-cooled fast reactor with power output of 10 or 50 MWe 30 years continuous operation without refueling

Safety design utilizing natural phenomena (automatic reactor shutdown and decay heat removal without human operation)

Reduction of maintenance introducing passive components

Enhanced safety and security by locating reactor building under ground level

Steam generator

Turbine generator

Figure 19.6 Schematic view of 4S reactor concept (taken from Tsuboi et al., 2009).

The 4S is a pool-type sodium-cooled uranium metallic fuel fast reactor, setting the core, the primary electro-magnetic pumps and the intermediate heat exchanger (IHX) inside the reactor vessel. The power output is 10 or 50 MWe. The average discharge burnup is about 35 GWd/t and the refueling interval is about 30 years. The core is narrow relative to its height (3.5 m) and is controlled by the movable six-segmented cylindrical reflector surrounding the core. The core can be shutdown by fall-down of the reflector to below the core region by the gravity. The reflector is able to shut the core down even if, for some reason, one of the six reflector segments is stuck.

As shown in Figure 19.6, the reactor vessel is located below ground level, providing substantial protection against an aircraft impact, and hence enhancing the security of the design. The containment system consists of the top dome and the guard vessel, which surrounds the rector vessel and the reflector drive equipment. The core heat

is transferred from the primary loop to the single intermediate heat transport system, and is then exchanged in the steam generator, also located below ground, to produce the steam, which drives the conventional steam turbine generator equipment in the water-steam loop. The 4S power plant can be constructed in a small site. The overall area covering the below-ground and above-ground structures is about 50 m long by 30 m wide.

The 4S has some passive safety features, such as the redundant residual heat removal system using only the natural circulation and the metallic core with the negative reactivity coefficients. It has two independent and redundant residual heat removal systems. They consist of the intermediate reactor auxiliary cooling system (IRACS), which removes decay heat using the air cooler installed in the intermediate heat transport system, and the reactor vessel auxiliary cooling system (RVACS), which removes heat transferred radially from the sodium coolant to the annulus between the guard vessel and a cylindrical steel heat collector located inside the cylindrical underground concrete wall.